First wall studies of a laser-fusion hybrid reactor design

International Nuclear Information System (INIS)

Hovingh, J.

1976-09-01

The design of a first wall for a 20 MW thermonuclear power laser fusion hybrid reactor is presented. The 20 mm thick graphite first wall is located 3.5 m from the DT microexplosion with a thermonuclear yield of 10 MJ. Estimates of the energy deposition, temperature, stresses, and material vaporized from the first wall due to the interaction of the x-rays, charged particle debris, and reflected laser light with the graphite are presented, along with a brief description of the analytical methods used for these estimations. Graphite is a viable first wall material for inertially-confined fusion reactors, with lifetimes of a year possible

Qualification Test for Korean Mockups of ITER Blanket First Wall

International Nuclear Information System (INIS)

Kim, S. K.; Lee, D. W.; Bae, Y. D.; Hong, B. G.; Jung, H. K.; Jung, Y. I.; Park, J. Y.; Jeong, Y. H.; Choi, B. K.; Kim, B. Y.

2009-01-01

ITER First Wall (FW) includes the beryllium armor tiles joined to CuCrZr heat sink with stainless steel cooling tubes. This first wall panels are one of the critical components in the ITER machine with the surface heat flux of 0.5 MW/m 2 or above. So qualification program needs to be performed with the goal to qualify the joining technologies required for the ITER First Wall. Based on the results of tests, the acceptance of the developed joining technologies will be established. The results of this qualification test will affect the final selection of the manufacturers for the ITER First Wall

Simulation of first-wall radiation effects

International Nuclear Information System (INIS)

Logan, C.M.; Anderson, J.D.; Hansen, L.F.

1975-01-01

Many of the effects induced in metals as a result of exposure to a radiation environment are intimately associated with the energy of primary recoil atoms (PKAs). Protons with an energy of 16 MeV closely reproduce the PKA energy spectrum which will be present at the first wall in a D--T fusion reactor and should therefore closely reproduce the radiation effects induced by PKAs in the first wall. A preliminary experiment with protons was conducted to measure the sputtering rate and to look for the phenomenon of chunk emission recently observed by Kaminsky and co-workers in samples exposed to 14-MeV neutrons. We are also able to observe the average projected transport range of activated PKAs. (U.S.)

Fail-safe first wall for preclusion of little leakage

International Nuclear Information System (INIS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

Leakages although excluded by design measures would occur most probably in highly stressed areas, weldments and locations without possibility to classify the state by in-service inspection. In a water-cooled first wall, allowable leak rate of water is generally very small, and therefore, locating of the leak portion under highly activated environment will be very difficult and be time-consuming. The double-wall concept is promising for the ITER first wall, because it can be made fail-safe by the application of the leak-before-break and the multiple load path concepts, and because it has a potential capability to solve the little leak problem. When the fail safe strength is well defined, subcritical crack growth in the damaged wall can be permitted. This will enable to detect stable leakage of coolant without deteriorating plasma operation. The paper deals with the little leak problem and presents method for evaluating small leak rate of a liquid coolant from crack-like defects. The fail-safe first wall with the double-wall concept is also proposed for preclusion of little leakage and its fail-safety is discussed. (author)

Quick installation/removal technology for first wall

International Nuclear Information System (INIS)

Tachikawa, Katsuhiro; Horie, Tomoyoshi; Seki, Yasushi; Fujisawa, Noboru; Kondoh, Mitsunori; Uchida, Takao.

1989-07-01

Fusion Next Step Device (FER) plans to experiment Deutrium-Tritium (D-T) reaction, remote handling and other fusion engineering issues. The fast neutron of 14 MeV caused by D-T reaction does not only activate the structural components inside the vacuum vessel, but also damages some first walls. The technique to remove the armour tiles of first walls by simple and quick operation is a key technology for the D-T burning Next Step Device. To establish the rational remote tile handling technology, consideration of consistency between the reactor structure and remote equipments should be made. The report comprises mainly the joint structures of armour tiles, design conditions (electro-magnetic force, cooling systems and so forth) and remote equipments. In addition, it is referred in shape memory alloy (SMA) applications, transportation of damaged tiles from the vacuum vessel and inspection systems for the first wall integrity. Hereafter, furthermore study in depth for the tile handling must be made in parallel with verification of remote systems and tile attachment structures using partial mockups. (author)

First wall thermal--mechanical analyses of the reference theta-pinch reactor

International Nuclear Information System (INIS)

Krakowski, R.A.; Hagenson, R.L.; Cort, G.E.

1977-01-01

The thermal-mechanical response of the Reference Theta-Pinch Reactor (RTPR) first wall was analyzed. The first wall problems anticipated for a pulsed, high-β fusion power plant can be ameliorated by either alterations in the physics operating point, materials reengineering, or blanket/first wall reconfiguration. Within the latter ''configuration'' scenario, a two-fold approach has been adopted for the thermal-mechanical portion of the RTPR first wall technology assessment. First, a number of new first wall configurations (bonded or unbonded laminated composites, all-ceramic structures, protective and/or sacrificial ''bumpers'') were considered. Second, a more quantitative failure criterion, based on the developing theories of fracture mechanics, was identified. For each first wall configuration, transient heat transfer and thermoelastic stress calculations have been made. Two-dimensional finite element structural analyses have been made for a variety of mechanical boundary conditions. Only the Al 2 O 3 /Nb - 1 Zr system has been considered. The results of this study indicated a wide range of design solutions to the pulsed thermal stress problem anticipated for the RTPR

ORNL facilities for testing first-wall components

International Nuclear Information System (INIS)

Tsai, C.C.; Becraft, W.R.; Gardner, W.L.; Haselton, H.H.; Hoffman, D.J.; Menon, M.M.; Stirling, W.L.

1985-01-01

Future long-impulse magnetic fusion devices will have operating characteristics similar to those described in the design studies of the Tokamak Fusion Core Experiment (TFCX), the Fusion Engineering Device (FED), and the International Tokamak Reactor (INTOR). Their first-wall components (pumped limiters, divertor plates, and rf waveguide launchers with Faraday shields) will be subjected to intense bombardment by energetic particles exhausted from the plasma, including fusion products. These particles are expected to have particle energies of approx.100 eV, particle fluxes of approx.10 18 cm -2 .s -1 , and heat fluxes of approx.1 kW/cm 2 CW to approx.100 kW/cm 2 transient. No components are available to simultaneously handle these particle and heat fluxes, survive the resulting sputtering erosion, and remove exhaust gas without degrading plasma quality. Critical issues for research and development of first-wall components have been identified in the INTOR Activity. Test facilities are needed to qualify candidate materials and develop components. At Oak Ridge National Laboratory (ORNL), existing neutral beam and wave heating test facilities can be modified to simulate first-wall environments with heat fluxes up to 30 kW/cm 2 , particle fluxes of approx.10 18 cm -2 .s -1 , and pulse lengths up to 30 s, within test volumes up to approx.100 L. The characteristics of these test facilities are described, with particular attention to the areas of particle flux, heat flux, particle energy, pulse length, and duty cycle, and the potential applications of these facilities for first-wall component development are discussed

Copper alloy conducting first wall for the FED-A tokamak

International Nuclear Information System (INIS)

Wiffen, F.W.

1984-01-01

The first wall of the tokamak FED-A device was designed to satisfy two conflicting requirements. They are a low electrical resistance to give a long eddy-current decay time and a high neutron transparency to give a favorable tritium breeding ratio. The tradeoff between these conflicting requirements resulted in a copper alloy first wall that satisfied the specific goals for FED-A, i.e., a minimum eddy-current decay time of 0.5 sec and a tritium breeding ratio of at least 1.2. Aluminum alloys come close to meeting the requirements and would also probably work. Stainless steel will not work in this application because shells thin enough to satisfy temperature and stress limits are not thick enough to give a long eddy-current decay time and to avoid disruption induced melting. The baseline first wall design is a rib-stiffened, double-wall construction. The total wall thickness is 1.5 cm, including a water coolant thickness of 0.5 cm. The first wall is divided into twelve 30-degree sectors. Flange rings at the ends of each sector are bolted together to form the torus. Structural support is provided at the top center of each sector

Reduced activation calculations for the STARFIRE first wall

International Nuclear Information System (INIS)

Mann, F.M.

1983-10-01

The activation of 27 elements (Li, Be, B, C, N, O, Mg, Al, Si, P, S, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Nb, Mo, Sn, Hf, Ta, W, and Pb) was calculated for a two-year exposure at the STARFIRE first-wall position. Based on a reasonable extension of current NRC regulations for near-surface land disposal, restrictions on N, Al, Ni, Cu, Nb, Mo, and Pb concentrations in first-wall materials may be required

INTOR impurity control and first wall system

International Nuclear Information System (INIS)

Abdou, M.A.

1983-04-01

The highlights of the recent INTOR effort on examining the key issues of the impurity control/first wall system are summarized. The emphasis of the work was an integrated study of the edge-region physics, plasma-wall interaction, materials, engineering and magnetic considerations associated with the poloidal divertor and pump limiter. The development of limiter and divertor collector plate designs with an acceptable lifetime was a major part of the work

Thermomechanical effects in a laser IFE first wall

International Nuclear Information System (INIS)

Blanchard, James P.; Martin, Carl J.

2005-01-01

Laser fusion chamber walls will experience large, pulsed heat loads at frequencies of several hertz. The heating, consisting of X-rays, neutrons, and ions, occurs over a few microseconds and is deposited volumetrically over the first few microns of the wall. For a reasonable chamber radius, the heating will be such that the surface temperature is a significant fraction of the melt temperature of the wall, and significant plasticity can be expected in ductile wall materials. This paper presents results for the transient temperatures and stresses in a tungsten-coated steel first wall for a laser fusion device. Failure analyses are carried out using both fatigue and fracture mechanics methodologies. The simulations predict that surface cracks are expected in the tungsten, but the cracks will arrest before reaching the substrate if the crack spacing is sufficiently small. In addition, the thermal and stress fields are compared for a laser fusion device with several simulation experiments. It is shown that the simulations can reproduce the peak surface temperatures, but the corresponding spatial distributions of the stress and temperature will be shallower than the reactor case

First wall response to energy disposition in conceptual laser fusion reactors

International Nuclear Information System (INIS)

Hovingh, J.

1976-02-01

Discussed are energy depositions in the first wall of various proposed laser-fusion reactors and the effect of pulse time on the stress and temperature in the first wall. Simple models can be used to estimate the temperature and stress rise from x-rays and neutrons. More complex analysis is needed to estimate the response of the first wall to reflected laser light and the pellet debris

Consolidation of HIP bonding technologies for the ITER first wall panels

International Nuclear Information System (INIS)

Sherlock, P.; Peacock, A.; Roedig, M.

2006-01-01

Over the last decade alternative technologies for the manufacture of the ITER first wall have been progressively developed. Now, as the build of ITER approaches, the manufacturing route is being consolidated around the best solutions found to date. The design of the first wall is based on the concept of blanket modules, each faced by separable first wall panels. For the manufacture of the first wall panels two HIP bonding technologies are proposed by AMEC NNC; the first to bond together the composite copper alloy / stainless steel heat sink base, the second to bond the beryllium tiles to the copper alloy surface of the heat sink base. These technologies have been developed incrementally through the use of experiments, part scale mock-ups and full scale first wall panel prototypes. This paper reviews the development of the HIP bonding technologies identified above and discusses the latest results from components produced by AMEC NNC under the auspices of EFDA. The manufacturing stages, non-destructive examination and heat flux test results from the work are presented for the latest first wall mock-up components. Conclusions are then drawn with regard to the important aspects for the series production of components for ITER. (author)

Conceptual design of the INTOR first-wall system

International Nuclear Information System (INIS)

Smith, D.L.; Majumdar, S.; Mattas, R.F.; Turner, L.; Jung, J.; Abdou, M.A.; Bowers, D.; Trachsel, C.; Merrill, B.

1981-10-01

The design concept and performance characteristics of the first-wall design for the phase-1 INTOR (International Tokamak Reactor) study is described. The reference design consists of a water-cooled stainless steel panel. The major uncertainty regarding the performance of the bare stainless steel wall relates to the response of a thin-melt layer predicted to form on limited regions during a plasma disruption. A more-complex backup design, which incorporates radiatively cooled graphite tiles on the inboard wall, is briefly described

Fabrication of a first wall panel by diffusion bonding

International Nuclear Information System (INIS)

Moreschi, L.F.; Pizzuto, A.; Alessandrini, I.

2002-01-01

Separated First Wall Panels mechanically attached to a shield block is now the reference concept for the Primary Wall Modules of RTO/RC ITER. The objective of the present work is to demonstrate the practical feasibility of a First Wall Panel utilizing a duplex round (steel) in square (copper) heat sink wound around a steel core and covered by Beryllium armour tiles. These three different materials (Be, Cu, steel) are joined together by diffusion bonding. The Copper alloy/stainless steel and Copper alloy/Beryllium joints were studied and developed selecting the optimal parameters for the related diffusion process. Several specimens were manufactured to be mechanically and thermally tested. The joints were mechanically tested using dedicated press equipment and investigated by micro-structural analysis with optical and SEM microscopy. Some thermal tests were finally carried out using an Electron Beam Facility. A dedicated R and D programme has led to the development of a co-drawing process, suitable for manufacturing the duplex Copper alloy-stainless steel heat sink. Two mock-ups were manufactured, the first in reduced-scale to test the thermal performance of the system, the second of larger scale and geometry better to represent the First Wall Panel

Thermal responses of tokamak reactor first walls during cyclic plasma burns

International Nuclear Information System (INIS)

Smith, D.L.; Charak, I.

1978-01-01

The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer code has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wall and liner have been determined for several burn-cycle scenarios and the extent of surface ablation that results from a plasma disruption has been determined for stainless steel and graphite first surfaces

Thermal responses of tokamak reactor first walls during cyclic plasma burns

International Nuclear Information System (INIS)

Smith, D.L.; Charak, I.

1977-01-01

The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wall and liner have been determined for several burn-cycle scenarios and the extent of surface ablation that results from a plasma disruption has been determined for stainless steel and graphite first surfaces

The mechanical performance of the fusion reactor first wall. Pt. 2

International Nuclear Information System (INIS)

Daenner, W.; Raeder, J.

1977-03-01

While the first part of this report was concerned with the steady-state mechanical analysis of the fusion reactor first wall, this part deals with the analysis based upon pulsed load conditions. In a first section we elaborate various solutions of the non-stationary heat conduction problem in plane geometry capable of describing the temperature response of the wall due to characteristic plasma pulse sequences. these solutions are input to a quasi-steady-state stress and strain analysis. Finally, the results of this analysis are set in relation to the fatigue properties of the wall material. A further section presents a description of a computer program which uses the mathematical procedure described. The results of some test runs are followed by those of detailed parameter studies. In the course of these calculations the influences of a number of design and operational quantities of a fusion reactor were investigated. It turned out that the choice of wall thickness and wall loading are of predominant importance for the first wall fatigue life. (orig.) [de

Fabrication of the full scale separable first wall of ITER shielding blanket

International Nuclear Information System (INIS)

Kosaku, Yasuo; Kuroda, Toshimasa; Hatano, Toshihisa; Enoeda, Mikio; Miki, Nobuharu; Akiba, Masato

2002-10-01

Shielding blanket for ITER-FEAT applies the unique first wall structure which is separable from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. Such unique features of blanket structure required technological clarification from the technical base of the previous achievement of the blanket module fabrication development. Previously, within the EDA Task T216+, a prototype for the no.4 Primary Wall Module of the ITER Shield Blanket with integrated first wall has been manufactured by forging and drilling and the first wall has been manufactured and joined to the shield block by Hot Isostatic Pressing (HIP) in one step process. This work has been performed to clarify the remaining R and D issues which have not been covered in the previous R and D. This report summarizes the demonstrative fabrication of the real scale separable first wall for ITER shielding blanket designed for ITER-FEAT, together with the essential technology developments such as, the slit grooving of the first wall with beryllium armor and SS shield block and fabrication of a partial mockup of beryllium armored first wall panel with built-in cooling channels. This work has been performed under the task agreement of G 16 TT 95 FJ (T420-1) in ITER Engineering Design Activity Extension Period. By the demonstration of the Be armor joining to the first wall panel, the joining technique of Be and DSCu developed previously, was shown to be applicable to the realistic structure of first wall panel. Also, the slit grooving by an end-mill method and an electron discharge machining method have been applied to the first wall mockup with Be armor tiles and demonstrated the applicability within the design tolerance. As the slit grooving technique

Development of fusion first-wall radiation damage facilities

International Nuclear Information System (INIS)

McElroy, R.J.; Atkins, T.

1986-11-01

The report describes work performed on the development of fusion-reactor first-wall simulation facilities on the Variable Energy Cyclotron, at Harwell, United Kingdom. Two irradiation facilities have been constructed: i) a device for helium and hydrogen filling up to 1000 ppm for post-irradiation mechanical properties studies, and ii) a helium implantation and damage facility for simultaneous injection of helium and radiation damage into a specimen under stress. These facilities are now fully commissioned and are available for investigations of first-wall radiation damage and for intercorrelation of fission- and fusion -reactor materials behaviour. (U.K.)

Development of real time monitoring for ITER first wall erosion

International Nuclear Information System (INIS)

Berryman, Ian.; Pallaras, Luke; Thomson, Laura; Wang, Michael; Riley, Daniel P.

2009-01-01

Full text: This project aims to contribute to the current research on the first wall erosion diagnostic for the ITER fusion reactor. The plasma-facing first wall tiles of the ITER tokamak reactor are exposed to an expected neutron flux of O. 7 8 M W/m2 and a thermal load of O. 5M W/m 2 during operation. Instabilities in the magnetically confined plasma, such as edge-Iocalised modes, cause the plasma to come into direct contact with the first wall. The resulting thermal loads can vaporise and ablate the tile material. Moreover, a flux of high-energy neutrons produced during the fusion process results in a range of radiation effects. Therefore, a diagnostic is necessary to monitor the extent and rate of damage caused to the first wall. We have considered and critically assessed the viability of six alternative diagnostic methods, encompassing both established and novel concepts. From these, a design featuring embedded conducting elements was selected as the strongest candidate, as by monitoring electrical signals it has the potential to detect both bulk erosion and radiation damage.

Diagnostic integration solutions in the ITER first wall

International Nuclear Information System (INIS)

Martínez, Gonzalo; Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger; Shigin, Pavel; Sabourin, Flavien; Gicquel, Stefan; Mitteau, Raphael; González, Jorge

2015-01-01

Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

Diagnostic integration solutions in the ITER first wall

Energy Technology Data Exchange (ETDEWEB)

Martínez, Gonzalo, E-mail: gonzalo.martinez@iter.org [Technical University of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Shigin, Pavel [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); National Research Nuclear University (MEPhI), Kashirskoe shosse, 115409 Moscow (Russian Federation); Sabourin, Flavien [ABMI-Groupe, Parc du Relais BatD 201 Route de SEDS, 13127 Vitrolles (France); Gicquel, Stefan; Mitteau, Raphael [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); González, Jorge [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain)

2015-10-15

Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

Modeling the thermodynamic response of metallic first walls

International Nuclear Information System (INIS)

Merrill, B.J.; Jones, J.L.

1982-01-01

The first wall material of a fusion device must have a high resistance to the erosion resulting from plasma disruptions. This erosion is a consequence of melting and surface vaporization produced by the energy deposition of the disrupting plasma. Predicting the extent of erosion has been the subject of various investigations, and as a result, the thermal modeling has evolved to include material melting, kinetics of surface evaporation, vaporized material transport, and plasma-vaporized material interactions. The significance of plasma-vapor interaction has yet to be fully resolved. The model presented by Hassanein suggests that the vapor attenuates the plasma ions, thereby shielding the wall surface and reducing the extent of vaporization. The erosion model developed by EG and G Idaho, Inc., has been extended to include a detailed model for plasma-vaporized material interaction. This paper presents the model, as well as predictions for plasma, vaporized material and first wall conditions during a disruption

Review of melting and evaporation of fusion-reactor first walls

International Nuclear Information System (INIS)

Fillo, J.A.; Makowitz, H.

1981-01-01

The most severe thermal loading on the first wall will occur when the plasma becomes unstable resulting in a hard plasma disruption or at the end of a discharge when the plasma is dumped on the wall in a very short period of time. Hard plasma disruptions are of particular concern in future fusion reactors where the thermal energy of the plasma may reach values on the order of 300 MJ. Sufficiently high heating rates can occur to melt the first wall surface, and the temperature can increase resulting in vaporization. Thermal models are reviewed which treat these problems

Materials for heat flux components of the first wall in fusion reactors

International Nuclear Information System (INIS)

Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1985-08-01

Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

First wall lifetime of the near term fusion reactors

International Nuclear Information System (INIS)

Matera, R.; Botti, S.; Cerrai, G.

1985-01-01

A sensitivity analysis of the influence of the operating conditions and of the design parameters over the first wall lifetime was performed by means of the computer program smile. In the range of operating conditions typical of an experimental fusion reactor like NET/INTOR and for a type AISI 316 stainless steel structural material, fatigue damage and fatigue crack growth are the limiting failure mechanisms of the first wall. The analysis shows in graphical form the limits of the allowable range of operating conditions or of design parameters

A design of a first wall for a demo reactor

International Nuclear Information System (INIS)

Bond, A.; Bond, R.A.; Cooke, P.I.H.

1985-01-01

A design of a first wall for a Demonstration reactor is reported based on an analysis of heat trasnport, sputtering damage, blanket neutronics and vacuum characteristics. The design comprises replaceable tungsten tiles radiatively cooled to a copper substrate, which in turn is cooled by high pressure helium. The overall engineering design of the first wall is described together with a discussion of the factors influencing the choice of design and materials

Overpower transient in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-09-01

The overpower transient from a plasma power excursion. The overpower transient considered in this report results from a postulated linear increase of the plasma power from the nominal generated power to four times this nominal power in 30 s. The Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design will be cooled by a number of separate cooling systems. The most important cooling systems are: The first wall cooling system, the blanket cooling system, the divertor cooling system, and the shield cooling system. In this report, the thermal-hydraulic analysis of the above-mentioned overpower transient will be presented for the first wall cooling system of NET/ITER. During overpower transients, the fusion power will increase to less than four times the nominal power. For this reason, the overpower transient considered in this report is the worst case scenario. The analysis of the thermal-hydraulic system behaviour during the considered overpower transient has been performed for a coolant temperature of 333 K (60 C) in the first wall inlet manifolds and 433 K (160 C) in the first wall outlet manifolds. The analysis has been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analysis, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. (orig.)

Material options for a commercial fusion reactor first wall

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m 2 . A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW 2 , provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys

Recent developments in fusion first wall, blanket, and shield technology

International Nuclear Information System (INIS)

Nygren, R.E.

1983-01-01

This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

Performance limits for fusion first-wall structural materials

International Nuclear Information System (INIS)

Smith, D.L.; Majumdar, S.; Billone, M.; Mattas, R.

2000-01-01

Key features of fusion energy relate primarily to potential advantages associated with safety and environmental considerations and the near endless supply of fuel. However, high-performance fusion power systems will be required in order to be an economically competitive energy option. As in most energy systems, the operating limits of structural materials pose a primary constraint to the performance of fusion power systems. In the case of fusion power, the first-wall/blanket system will have a dominant impact on both economic and safety/environmental attractiveness. This paper presents an assessment of the influence of key candidate structural material properties on performance limits for fusion first-wall blanket applications. Key issues associated with interactions of the structural materials with the candidate coolant/breeder materials are discussed

Conception of thermonuclear reactor with a shielding layer of the first wall

International Nuclear Information System (INIS)

Marin, S.V.

1979-01-01

Considered is the way of the shielding of the first wall of a thermonuclear reactor by the layer of ISSEC (Internal spectral shifter and Energy Converter). It is a constructive non-power element placed between a plasma and the first wall, and intended for the softening of the spectrum and intensity reduction of particle fluxes falling on the first wall. Results of neutron-physical calculations of the UWMAK-type reactor blanket (in the S 4 -P 3 approximation) are presented. While comparing five materials (C, Mo, Nb, V,W) by the rate of radiation damage formation, gas production, radioactivity level and energy output in the blanket with the 316 stainless steel first wall, it is obvious that the conception of ISSEC permits to prolong the service period of the first wall. Construction elements should be then in the same irradiation conditions as those in fast reactors. Molybdenum has been taken as the best ISSEC material. It reduces the number of displaced atoms of the first wall by 20% and decreases helium production by about 100%, increases energy output in the blanket by 15-18%. However, graphite is advantageous, while comparing it to molybdenum in values of residual energy output, radioactivity level, costs and manufacture simplicity. One problem stays unsolved, which is connected with chemical sputtering of graphite at the formation of C 2 H 2 in the high temperature range. So it is hard to prefer any material now

Falling liquid film flow along cascade-typed first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, T.; Nakai, T.; Kawara, Z.

2007-01-01

To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a 'cascade-typed' first wall with a falling liquid film flow is proposed as the 'liquid wall' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the STREAM code and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ρu 2 δ/σ: ρ is density, u is velocity, δ is film thickness, σ is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant water-head located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same structure and the same height as the reactor design

Neutron-transparent first wall for module testing

International Nuclear Information System (INIS)

Fuller, G.M.; Cramer, B.A.; Haines, J.R.; Kirchner, J.; Engholm, B.A.; Seki, M.

1983-01-01

Major design goals for FED-R are the achievement of: (1) a high level of neutron exposure of the test modules and (2) a capability for rapid changeout of test modules. A major factor in rapid changeout is perceived to be the location of the vacuum boundary. In FED-R this boundary was set at the first wall so that module changeout did not require the plasma chamber to be brought up to atmosphere. Efforts to realize these goals in the design resulted in a neutronically thin outboard wall for the vacuum vessel constructed of 316 stainless steel (SS) with helium as a coolant. A normalized 14-MeV neutron transmission of 0.82 is expected, with an inlet pressure of 2 MPa and a pumping power requirement of 8.7 MW. Other options considered in the study were aluminum as a wall material and water and sodium potassium (NaK) as coolants

Lithium adsorption by the first wall of fusion reactor-tokamak

International Nuclear Information System (INIS)

Bakunin, O.G.

1989-01-01

Lithium adsorption by the first wall of fusion reactor under stationary conditions and in the absence of chemical reactions is considered. Possibility of achieving 70% coating of the wall with lithium which can lead to sufficient decrease of sputtering is shown. 5 refs.; 5 figs

First wall

International Nuclear Information System (INIS)

Omori, Junji.

1991-01-01

Graphite and C/C composite are used recently for the first wall of a thermonuclear device since materials with small atom number have great impurity allowable capacity for plasmas. Among them, those materials having high thermal conduction are generally anisotropic and have an upper limit for the thickness upon production. Then, anisotropic materials are used for a heat receiving plate, such that the surfaces of the heat receiving plate on the side of lower heat conductivity are brought into contact with each other, and the side of higher thermal conductivity is arranged in parallel with small radius direction and the toroidal direction of the thermonuclear device. As a result, the incident heat on an edge portion can be transferred rapidly to the heat receiving plate, which can suppress the temperature elevation at the surface to thereby reduce the amount of abrasion. Since the heat expansion coefficient of the anisotropic materials is great in the direction of the lower heat conductivity and small in the direction of the higher heat conductivity, the gradient of a thermal load distribution in the direction of the higher heat expansion coefficient is small, and occurrence of thermal stresses due to temperature difference is reduced, to improve the reliability. (N.H.)

Some stress-related issues in tokamak fusion reactor first walls

International Nuclear Information System (INIS)

Majumdar, S.; Pai, B.; Ryder, R.H.

1987-01-01

Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

Thermal stress and creep fatigue limitations in first wall design

International Nuclear Information System (INIS)

Majumdar, S.; Misra, B.; Harkness, S.D.

1977-01-01

The thermal-hydraulic performance of a lithium cooled cylindrical first wall module has been analyzed as a function of the incident neutron wall loading. Three criteria were established for the purpose of defining the maximum wall loading allowable for modules constructed of Type 316 stainless steel and a vanadium alloy. Of the three, the maximum structural temperature criterion of 750 0 C for vanadium resulted in the limiting wall loading value of 7 MW/m 2 . The second criterion limited thermal stress levels to the yield strength of the alloy. This led to the lowest wall loading value for the Type 316 stainless steel wall (1.7 MW/m 2 ). The third criterion required that the creep-fatigue characteristics of the module allow a lifetime of 10 MW-yr/m 2 . At wall temperatures of 600 0 C, this lifetime could be achieved in a stainless steel module for wall loadings less than 3.2 MW/m 2 , while the same lifetime could be achieved for much higher wall loadings in a vanadium module

Irradiation creep lifetime analysis on first wall structure materials for CFETR

Energy Technology Data Exchange (ETDEWEB)

Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

2017-05-15

Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

Role of inert gases in first wall phenomena in fusion devices

International Nuclear Information System (INIS)

Das, S.K.

1979-01-01

The first wall surfaces of fusion devices will be exposed to bombardment by inert gaseous projectiles such as helium. The flux, energy and angular distribution of the helium radiation will depend not only on the type of device but also on its design parameters. For near term tokamak devices, the first wall surface phenomena caused by helium bombardment that appear to be quite important are physical sputtering and radiation blistering. Examples of these processes for a number of first wall candidate materials are discussed. While the physical sputtering phenomen is well understood, the mechanism of blister formation is still not fully understood. The various models proposed for radiation blistering of metal during helium bombardment is critically reviewed in the light of most recent experimental results

First wall fusion blanket temperature variation - slab geometry

International Nuclear Information System (INIS)

Fillo, J.A.

1978-01-01

The first wall of a fusion blanket is approximated by a slab, with the surface facing the plasma subjected to an applied heat flux, while the rear surface is convectively cooled. The relevant parameters affecting the heat transfer during the early phases of heating as well as for large times are established. Analytical solutions for the temperature variation with time and space are derived. Numerical calculations for an aluminum and stainless steel slab are performed for a wall loading of 1 MW(th)/m 2 . Both helium and water cooling are considered. (Auth.)

First wall thermomechanical stress analysis in a fusion ignition experiment

International Nuclear Information System (INIS)

Rodin, G.; Carrera, R.; Howell, J.; Hwang, Y.L.; Montalvo, E.; Ordonez, C.; Dong, J.Q.

1990-01-01

The fusion ignition experiment IGNITEX + has been proposed as a low cost means of producing and controlling fusion ignited plasmas for scientific study. A single-turn-coil tokamak plasmas for scientific study. A single-turn-coil tokamak cryogenically precooled at liquid nitrogen temperature is used to produce 20 T fields and 12 MA plasma currents so that high-density ohmic ignition is possible. The high-field, high-density operation should maintain the plasma relatively free of wall impurities. In order to minimize plasma cooling, a low-Z first wall is considered for IGNITEX. The IGNITEX design philosophy emphasizes simplicity and low cost. A limiterless, smooth first will without files and plates is proposed. A low-Z material is applied by plasma jet techniques over a resistive vacuum vessel. This design is thought to be adequate for a magnetic fusion ignition experiment. Maintenance and operation of the first wall system is significantly simplified when compared to conventional designs

A three-bar model for ratcheting of fusion reactor first wall

International Nuclear Information System (INIS)

Wolters, J.; Majumdar, S.

1994-12-01

First wall structures of fusion reactors are subjected to cyclic bending stresses caused by inhomogeneous temperature distribution during plasma burn cycles and by electromagnetically induced impact loads during plasma disruptions. Such a combination of loading can potentially lead to ratcheting or incremental accumulation of plastic strain with cycles. An elastic-plastic three-bar model is developed to investigate the ratcheting behavior of the first wall

First Wall and Operational Diagnostics

International Nuclear Information System (INIS)

Lasnier, C; Allen, S; Boedo, J; Groth, M; Brooks, N; McLean, A; LaBombard, B; Sharpe, J; Skinner, C; Whyte, D; Rudakov, D; West, W; Wong, C

2006-01-01

In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. There are sections discussing infrared cameras, visible and VUV cameras, pressure gauges and RGAs, Langmuir probes, thermocouples, and erosion and deposition measurements by insertable probes and quartz microbalance. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and IR cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER

Methodology for first wall design

International Nuclear Information System (INIS)

Galambos, J.D.; Conner, D.L.; Goranson, P.L.; Lousteau, D.C.; Williamson, D.E.; Nelson, B.E.; Davis, F.C.

1993-01-01

An analytic parametric scoping tool has been developed for application to first wall (FW) design problems. Both thermal and disruption force effects are considered. For the high heat flux and high disruption load conditions expected in the International Thermonuclear Experimental Reactor (ITER) device, Vanadium alloy and dispersion-strengthened copper offer the best stress margins using a somewhat flattened plasma-facing configuration. Ferritic steels also appear to have an acceptable stress margin, whereas the conventional stainless steel 316 does not appear feasible. If a full semicircle shape FW is required, only the Vanadium and ferritic steel alloy have acceptable solutions

Implantation measurements to determine tritium permeation in first wall structures

International Nuclear Information System (INIS)

Holland, D.F.; Causey, R.A.

1983-01-01

A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium atoms in a stainless steel disk to simulate tritium implantation in first wall structures. The permeation rate was measured under various operating conditions. These results were used in the TMAP computer code to determine potential tritium loss rates for fusion reactors

Implantation measurements to determine tritium permeation in first-wall structures

International Nuclear Information System (INIS)

Holland, D.F.; Causey, R.A.; Sattler, M.L.

1983-01-01

A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first-wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium atoms in a stainless steel disk to simulate tritium implantation in first-wall structures. The permeation rate was measured under various operating conditions. These results were used in the TMAP computer code to determine potential tritium loss rates for fusion reactors

Low-Z coating as a first wall of nuclear fusion devices

International Nuclear Information System (INIS)

Shikama, Tatsuo; Okada, Masatoshi

1984-01-01

The tokamak nuclear fusion devices of the largest scale in the world, TFTR in USA and JET in Europe, started the operation from the end of 1982 to 1983. Also in Japan, the tokamak JT-60 is scheduled to begin the operation in 1985. One of the technological obstacles is the problem of first walls facing directly to plasma and subjected to high particle loading and thermal loading. Moreover, first walls achieve the active role of controlling impurities in plasma and recycling hydrogen isotopes. It is impossible to find a single material which satisfies all these requirements. The compounding of materials can create a material having new function, but also has the meaning of expanding the range of material selection. One of the material compounding methods is surface coating. In this paper, as the materials for first walls, the characteristics of low Z materials are discussed from the design examples of actual takamak nuclear fusion devices. The outline of first walls is explained. High priority is given to the impurity control in plasma, and in view of plasma energy emissivity and the rate of self sputtering, low Z material coating seems to be the solution. The merits and the problems of such low Z material coating are discussed. (Kako, I.)

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

International Nuclear Information System (INIS)

Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

2013-01-01

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m 2 . It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

Energy Technology Data Exchange (ETDEWEB)

Igitkhanov, Yu., E-mail: juri.igitkhanov@lhm.fzk.de [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Bazylev, B.; Landman, I. [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Boccaccini, L. [Karlsruhe Institute of Technology, INR, Karlsruhe (Germany)

2013-07-15

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m{sup 2}. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

Science.gov (United States)

Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

2013-07-01

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ˜14 MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

Frost as a first wall for the ICF laboratory microfusion facility

International Nuclear Information System (INIS)

Orth, C.D.

1989-01-01

The authors introduce the concept of using frost as the first wall of the ICF Laboratory Microfusion Facility being designed to produce 200-1000 MJ of thermonuclear yield. They present one design incorporating 2cm of frost deposited at 0.1 g/cm/sup 3/ on an LN-cooled fiber-reinforced polymer substrate. They calculate that such a frost layer will protect the substrate from ablation by target x rays and debris, and from shock-induced spallation. Postshot washdown with water should permit low-activation operation, and should preserve the original wall properties. The authors expect the impact of the frost on laser optics to be minimal, and expect the preshot lifetime of thermally unprotected cryogenic targets to be extended by operating the wall at 100-150 K. Moreover, they believe that such a frost first wall involves little technical risk, and will be inexpensive to construct and operate

Frost as a first wall for the ICF Laboratory Microfusion Facility

International Nuclear Information System (INIS)

Orth, C.D.

1988-01-01

We introduce the concept of using frost as the first wall of the ICF Laboratory Microfusion Facility being designed to produce 200--1000 MJ of thermonuclear yield. We present one design incorporating 2 cm of frost deposited at 0.1 g/cm 3 on an LN-cooled fiber-reinforced polymer substrate. We calculate that such a frost layer will protect the substrate from ablation by target x rays and debris, and from shock-induced spallation. Postshot washdown with water should permit low-activation operation, and should preserve the original wall properties. We expect the impact of the frost on laser optics to be minimal, and expect the preshot lifetime of thermally unprotected cryogenic targets to be extended by operating the wall at 100-150 K. Moreover, we believe that such a frost first wall will involve little technical risk, and will be inexpensive to construct and operate. 4 refs., 1 fig

Investigation of cascade-type falling liquid-film along first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, T.; Nakai, T.; Kawara, Z.; Norimatsu, T.; Kozaki, Y.

2008-01-01

To protect the first wall of an inertia fusion reactor from extremely high heat flux, X-rays, alpha particles and fuel debris caused by a nuclear fusion reaction, a 'cascade-type' falling liquid-film flow is proposed as a 'liquid-wall' concept. The flow visualization experiment to investigate the feasibility of this liquid-wall concept has been conducted. The preliminary numerical simulation results suggest that the current cascade structure design should be improved because less thermal-mixing is expected. The cascade-type structure has, therefore, been redesigned. This new cascade-type first wall consists of a liquid reservoir which has a free-surface to maintain a constant water head in the rear, and connects to a slit composed of two plates, i.e., the first wall is connected to a slit which is partially made up of the first wall to begin with it. The numerical simulations were performed on the new cascade-type first wall and they show the stable liquid-film flow on it. Moreover, the POP (proof-of-principle) flow visualization experiments, which satisfy the Weber number coincident condition, are carried out using water as the working fluid. By comparing the numerical and experimental results, it was found that the liquid-film flow with 3-5 mm thickness could be stably established. According to these results for the new cascade-type first wall concept, it was confirmed that the coolant flow rate and the thickness of the liquid-film could be controlled if the Weber number coincident condition was satisfied

Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

International Nuclear Information System (INIS)

Kim, Byoung-Yoon; Ahn, Hee-Jae

2011-01-01

The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

Options for a high heat flux enabled helium cooled first wall for DEMO

Energy Technology Data Exchange (ETDEWEB)

Arbeiter, Frederik, E-mail: f.arbe@kit.edu; Chen, Yuming; Ghidersa, Bradut-Eugen; Klein, Christine; Neuberger, Heiko; Ruck, Sebastian; Schlindwein, Georg; Schwab, Florian; Weth, Axel von der

2017-06-15

Highlights: • Design challenges for helium cooled first wall reviewed and otimization approaches explored. • Application of enhanced heat transfer surfaces to the First Wall cooling channels. • Demonstrated a design point for 1 MW/m{sup 2} with temperatures <550 °C and acceptable stresses. • Feasibility of several manufacturing processes for ribbed surfaces is shown. - Abstract: Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of flexible temperature range, chemical inertness, no activation, comparatively low effort to remove tritium from the gas and no chemical corrosion. Existing blanket designs have shown the ability to use helium cooled first walls with heat flux densities of 0.5 MW/m{sup 2}. Average steady state heat loads coming from the plasma for current EU DEMO concepts are expected in the range of 0.3 MW/m{sup 2}. The definition of peak values is still ongoing and depends on the chosen first wall shape, magnetic configuration and assumptions on the fraction of radiated power and power fall off lengths in the scrape off layer of the plasma. Peak steady state values could reach and excess 1 MW/m{sup 2}. Higher short-term transient loads are expected. Design optimization approaches including heat transfer enhancement, local heat transfer tuning and shape optimization of the channel cross section are discussed. Design points to enable a helium cooled first wall capable to sustain heat flux densities of 1 MW/m{sup 2} at an average shell temperature lower than 500 °C are developed based on experimentally validated heat transfer coefficients of structured channel surfaces. The required pumping power is in the range of 3–5% of the collected thermal power. The FEM stress analyses show code-acceptable stress intensities. Several manufacturing methods enabling the application of the suggested heat transfer enhanced first wall channels are explored. An

Electromagnetic forces distribution and mechanical analysis in the first wall structure for INTOR/NET

International Nuclear Information System (INIS)

Coccorese, E.; Martone, R.; Rubinacci, G.; Biggio, M.; Inzaghi, A.; Turri, M.

1984-01-01

In the context of the studies performed at JRC-Ispra for NET/INTOR, a modular stainless steel first wall, and separated from the blanket which it envelops has been proposed. During plasma disruption the metallic structure of the first wall is inevitably subject to appreciable electromagnetic forces caused by induced eddy current-magnetic field interactions. The deformation and stress distributions in the first wall were quantified at various instants of time by three-dimensional calculations using the ICES-STRUDL code. (author)

Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2

International Nuclear Information System (INIS)

Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio

1997-01-01

A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)

Fabrication of prototype mockups of ITER shielding blanket with separable first wall

International Nuclear Information System (INIS)

Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Akiba, Masato

2002-07-01

Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

Magnetic forces on a ferromagnetic HT-9 first wall/blanket and coolant pipe

International Nuclear Information System (INIS)

Lechtenberg, T.A.; Dahms, C.; Attaya, H.; Univ. of Wisconsin, Madison)

1984-01-01

The GFUN 3D code was used to model the toroidal fields and determine the magnetic body forces on the STARFIRE design for coolant pipes exiting the first wall sector and first wall/blanket modules. The HT-9 coolant pipes were modeled on the basis of a square bar having the same length and material volume as the coolant pipes. The stress analysis was performed using these magnetic forces applied to a pipe of 4 meters length, 8.25 cm O.D., and 0.75 cm thickness by the MODSAP stress analysis code. For the first wall/blanket module, GFUN 3D does not allow full modeling of the complex thin-walled structure or numerous small tubes because of the element aspect ratio limitations. Therefore, to obtain three dimensional loads, a solid homogeneous equivalent structure was used

An electrically conducting first wall for the fusion engineering device-A (FED-A) tokamak

International Nuclear Information System (INIS)

Cramer, B.A.; Fuller, G.M.

1983-01-01

The first wall of the tokamak FED-A device was designed to satisfy two conflicting requirements. They are a low electrical resistance to give a long eddy-current decay time and a high neutron transparency to give a favorable tritium breeding ratio. The tradeoff between these conflicting requirements resulted in a copper alloy first wall that satisfied the specific goals for FED-A, i.e., a minimum eddy-current decay time of 0.5 sec and a tritium breeding ratio of at least 1.2. Aluminum alloys come close to meeting the requirements and would also probably work. Stainless steel will not work in this application because shells thin enough to satisfy temperature and stress limits are not thick enough to give a long eddy-current decay time and to avoid disruption induced melting. The baseline first wall design is a rib-stiffened, double-wall construction. The total wall thickness is 1.5 cm, including a water coolant thickness of 0.5 cm. The first wall is divided into twelve 30-degree sectors. Flange rings at the ends of each sector are bolted together to form the torus. Structural support is provided at the top center of each sector

First wall and blanket module safety enhancement by material selection and design decision

International Nuclear Information System (INIS)

Merrill, B.J.

1980-01-01

A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems

A first wall material probe manipulator for the 'TEXTOR' tokamak

International Nuclear Information System (INIS)

Marmy, P.; Stiefel, U.

1984-04-01

Textor is a technology oriented tokamak of Euratom at the Kernforschungsanlage Juelich (KFA). Switzerland participates in its experimental program within the framework of the IEA agreement on Plasma Wall Interaction. A major task of EIR consists in the layout, construction and fabrication of a manipulator for the remote handling of up to 240 specimen candidate first wall materials. This operation has to be done without breaking the ultra high vacuum (UHV) and with wall temperatures up to 300 0 C. A great number of preexperiments involving different materials had to be carried out; the understanding of the tribology in ultra high vacuum could be improved. (Auth.)

Deuterium implantation in first wall candidate materials by exposure in the Princeton large torus

Energy Technology Data Exchange (ETDEWEB)

Chang, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center); Manos, D. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

Titanium alloys are of interest as a first wall material in fusion reactors because of their excellent thermophysical and thermomechanical properties. A major concern with their application to the first wall is associated with the known affinity of titanium for hydrogen and the related consequences for fuel recycling, tritium inventory, and hydrogen embrittlement. Little information exists on trapping and release of hydrogen isotopes implanted at energies below 500 eV. This work was undertaken to measure hydrogen isotope trapping and release at the first wall of the Princeton Large Torus Tokamak (PLT).

Automation of fusion first wall design using artificial intelligence technique

International Nuclear Information System (INIS)

Yoshimura, Shinobu; Yagawa, Genki; Mochizuki, Yoshihiko

1990-01-01

This paper describes the application of artificial intelligence techniques to a design automation of the fusion first wall to be operated in the complex environment where huge electromagnetic and thermal loading as well as heavy neutron irradiation occur. As a basic strategy of designing structure shape considering many coupled phenomena, an ordinary design procedure based on the generate and test strategy is adopted because of its simplicity and broad applicability. To automate the design procedure with maintaining its flexibility, extensibility and efficiency, artificial intelligence techniques are utilized in the following. An object-oriented knowledge representation technique is adopted to store knowledge modules, that is, objects, related to the first wall design, while a data-flow processing technique is utilized as an inference mechanism among the knowledge modules. These techniques realize the flexibility and extensibility of the system. Moreover, as an efficient design modification mechanism, which is essential in a design process, an empirical approach based on experts' empirical knowledge and a mathematical approach based on a kind of numerical sensitivity analysis are introduced. The developed system is applied to a simple example of the design of a two-dimensional model of the first wall with a cooling channel, and its fundamental performance is clearly demonstrated. (author)

First-wall/blanket materials selection for STARFIRE tokamak reactor

International Nuclear Information System (INIS)

Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

1980-01-01

The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed

Control of first-wall surface conditions in the 2XIIB Magnetic Mirror Plasma Confinement experiment

International Nuclear Information System (INIS)

Simonen, T.C.; Bulmer, R.H.; Coensgen, F.H.

1976-01-01

The control of first-wall surface conditions in the 2XIIB Magnetic Mirror Plasma Confinement experiment is described. Before each plasma shot, the first wall is covered with a freshly gettered titanium surface. Up to 5 MW of neutral beam power has been injected into 2XIIB, resulting in first-wall bombardment fluxes of 10 17 atoms . cm -2 . s -1 of 13-keV mean energy deuterium atoms for several ms. The background gas flux is measured with a calibrated, 11-channel, fast-atom detector. Background gas levels are found to depend on surface conditions, injected beam current, and beam pulse duration. For our best operating conditions, an efective reflex coefficient of 0.3 can be inferred from the measurements. Experiments with long-duration and high-current beam injection are limited by charge exchange; however, experiments with shorter beam duration are not limited by first-wall surface conditions. It is concluded that surface effects will be reduced further with smoother walls. (Auth.)

Hydrogen isotope behavior in the first wall of JT-60U after deuterium plasma operation

International Nuclear Information System (INIS)

Oya, Y.; Tanabe, T.; Oyaidzu, M.; Shibahara, T.; Sugiyama, K.; Yoshikawa, A.; Onishi, Y.; Hirohata, Y.; Ishimoto, Y.; Yagyu, J.; Arai, T.; Masaki, K.; Okuno, K.; Miya, N.; Tanaka, S.

2007-01-01

Retention of hydrogen isotopes in the carbon (isotropic graphite) first wall tiles of JT-60U was studied by secondary ion mass spectrometry and thermal desorption spectroscopy. The surface morphology and erosion/deposition profiles of the tiles were characterized using scanning electron microscope and X-ray photoelectron spectroscopy. The upper area is mainly eroded, while the bottom area of the inboard wall is dominated by deposition. In contrast to the divertor area, hydrogen isotope retention in the eroded wall area was generally larger than that in the deposition dominated area. Measured near surface concentrations of hydrogen isotopes in the wall tiles, as well as the D/H ratios, were a little higher than those in the divertor area. This indicates direct implantation of high-energy D from NBI into the first wall. The lower temperature of the first wall relative to the divertor tiles would reduce desorption and/or replacement of implanted D by subsequent D or H impingement

In service experience feed back of the tore supra actively cooled inner first wall

International Nuclear Information System (INIS)

Schlosser, J.; Chappuis, P.; Chatelier, M.; Cordier, J.J.; Deschamps, P.; Garampon, L.; Guilhem, D.; Lipa, M.; Mitteau, R.

1994-01-01

Over 12000 plasma shots (some of them up to 8 MW of additional power and same as long as 60 s) have been achieved in TORE SUPRA (TS) with a significant number of them limited by thr inner first wall. This actively water cooled wall is covered with brazed graphite tiles. High power - high energy experiments have shown that a reliability of the graphite tile/heat sink joint and an accurate alignment of the wall are needed. This paper summarizes the experience gained with this component and developments in progress in order to improve the performance of such a inner first wall. (authors). 9 refs., 13 figs., 2 tabs

In service experience feed back of the tore supra actively cooled inner first wall

Energy Technology Data Exchange (ETDEWEB)

Schlosser, J; Chappuis, P; Chatelier, M; Cordier, J J; Deschamps, P; Garampon, L; Guilhem, D; Lipa, M; Mitteau, R

1994-12-31

Over 12000 plasma shots (some of them up to 8 MW of additional power and same as long as 60 s) have been achieved in TORE SUPRA (TS) with a significant number of them limited by thr inner first wall. This actively water cooled wall is covered with brazed graphite tiles. High power - high energy experiments have shown that a reliability of the graphite tile/heat sink joint and an accurate alignment of the wall are needed. This paper summarizes the experience gained with this component and developments in progress in order to improve the performance of such a inner first wall. (authors). 9 refs., 13 figs., 2 tabs.

Manufacturing routes for stainless steel first wall panels

International Nuclear Information System (INIS)

Bucci, Ph.; Federzoni, L.; Le Marois, G.; Lorenzetto, P.

2001-01-01

Hot isostatic pressing (HIP) techniques are being considered in the European Home Team as one of the fabrication routes to produce ITER-FEAT primary first wall panels (PFWP). To demonstrate the potential and the availability of such techniques, material development, innovative mock-up fabrications and numerical modeling for the production of near-net shape components are currently been studied by CEA/CEREM in collaboration with the EFDA-CSU Garching. The aim of this work is to investigate the manufacturing feasibility of advanced PFWP concepts, with reduced pitch between FW cooling channels and reduced material thickness between the FW cooling channels and the front surface, in order to improve the thermal fatigue performance of these concepts. In order to select the best fabrication route, two different manufacturing methods based on the HIP process are being considered. The first one consists in manufacturing of the first wall panel by a HIP forming technique. Mock-ups are made of a serpentine tube expanded into a proper matrix. 2-D computer modeling has been performed to estimate the serpentine deformation. The second manufacturing route is based on the powder HIP technique. Mock-ups have been made of a serpentine embedded into SS powder. In both cases, the objective was to obtain the minimum pitch between the stainless steel (SS) tubes and between the SS tubes and the front face

Investigation of cascade-typed falling liquid film flow along first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, Tomoaki; Nakai, Tadakatsu; Kawara, Zensaku

2007-01-01

To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a ''cascade-typed'' falling liquid film flow is proposed as the ''liquid wall'' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the commercial code (STREAM: unsteady three-dimensional general purpose thermofluid code) and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ru 2 d/s: r is density, u is velocity, d is film thickness, s is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant waterhead located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same

Surface condition effects on tritium permeation through the first wall of a water-cooled ceramic breeder blanket

Energy Technology Data Exchange (ETDEWEB)

Zhou, H.-S. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei (China); Xu, Y.-P.; Liu, H.-D. [Science Island Branch of Graduate School, University of Science and Technology of China, P.O. Box 1126, Hefei (China); Liu, F.; Li, X.-C.; Zhao, M.-Z.; Qi, Q.; Ding, F. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei (China); Luo, G.-N., E-mail: gnluo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei (China); Science Island Branch of Graduate School, University of Science and Technology of China, P.O. Box 1126, Hefei (China); Hefei Center for Physical Science and Technology, P.O. Box 1126, Hefei (China); Hefei Science Center of Chinese Academy of Science, P.O. Box 1126, Hefei (China)

2016-11-01

Highlights: • We investigate surface effects on T transport through the first wall. • We solve transport equations with various surface conditions. • The RAFMs walls w/and w/o W exhibit different T permeation behavior. • Diffusion in W has been found to be the rate-limiting step. - Abstract: Plasma-driven permeation of tritium (T) through the first wall of a water-cooled ceramic breeder (WCCB) blanket may raise safety and other issues. In the present work, surface effects on T transport through the first wall of a WCCB blanket have been investigated by theoretical calculation. Two types of wall structures, i.e., reduced activation ferritic/martensitic steels (RAFMs) walls with and without tungsten (W) armor, have been analyzed. Surface recombination is assumed to be the boundary condition for both the plasma-facing side and the coolant side. It has been found that surface conditions at both sides can affect T permeation flux and inventory. For the first wall using W as armor material, T permeation is not sensitive to the plasma-facing surface conditions. Contamination of the surfaces will lead to higher T inventory inside the first wall.

First nitrogen-seeding experiments in JET with the ITER-like Wall

Energy Technology Data Exchange (ETDEWEB)

Oberkofler, M., E-mail: martin.oberkofler@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Douai, D. [CEA Centre de Cadarache, 13108 Saint Paul lez Durance, Cedex (France); Brezinsek, S.; Coenen, J.W. [Institut für Energie- und Klimaforschung, IEK-4, TEC, Association EURATOM-FZJ, 52425 Jülich (Germany); Dittmar, T. [Center for Energy Research, University of California–San Diego, 9500 Gilman Dr., San Diego, CA 92093-0417 (United States); Drenik, A. [Jožef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia); Romanelli, S.G. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Joffrin, E. [CEA Centre de Cadarache, 13108 Saint Paul lez Durance, Cedex (France); McCormick, K. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Brix, M. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Calabro, G. [Associazione EURATOM-ENEA sulla Fusione, Via E. Fermi 45 FRASCATI-Roma (Italy); Clever, M. [Institut für Energie- und Klimaforschung, IEK-4, TEC, Association EURATOM-FZJ, 52425 Jülich (Germany); Giroud, C. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Kruezi, U. [Institut für Energie- und Klimaforschung, IEK-4, TEC, Association EURATOM-FZJ, 52425 Jülich (Germany); Lawson, K. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Linsmeier, Ch. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); and others

2013-07-15

In this contribution we present results from the first N{sub 2} seeding experiments in JET performed after installation of the ITER-like Wall. Gas balance measurements for seeded L-mode discharges indicate very strong N{sub 2} retention as well as a potential increase in D{sub 2} retention. The possible influence of ammonia production on this apparent retention is discussed. Plasma parameters and impurity content were monitored throughout the seeded discharges as well as during subsequent clean-up discharges. These experiments give first insight into phenomena related to the use of nitrogen as seeding gas in JET with the ITER-like Wall, such as ammonia production and nitrogen legacy.

Analysis of three loss-of-flow accidents in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-05-01

This report presents the thermal-hydraulic analysis of three Loss-of-Flow Accidents (LOFAs) in the first wall cooling system of the Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design. The LOFAs considered result from a loss of the forced coolant flow caused by a loss of electrical power for the recirculation pump in the primary circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. In the LOFA case without plasma shutdown, melting starts in the first wall about 150 s after accident initiation. In the LOFA case with delayed plasma shutdown, melting starts in the first wall when the plasma shutdown is initiated later than about 110 s after accident initiation. Melting does not occur in the first wall during a LOFA with prompt plasma scram. (orig.)

First wall and blanket stresses induced by cyclic fusion core operations

International Nuclear Information System (INIS)

Bohachevsky, I.O.; Kostoff, R.N.

1981-01-01

An analysis is made of cyclic thermal loads and stresses for the complete range of operating conditions. Two critical components were examined; the solid wall adjacent to the fusion plasma (first wall) and the fuel elements in the high power density region of the blanket. Simple closed form expressions were derived for temperature increases and thermal stresses that may be evaluated conveniently and rapidly and the values compared for different systems

Structural response of a Tokamak first wall under electromagnetic forces caused by a plasma disruption

International Nuclear Information System (INIS)

Crutzen, Y.R.; Biggio, M.; Farfaletti-Casali, F.; Antonacci, P.; Vitali, R.

1987-01-01

The modern computerized techniques of CAD/FEM analysis are extensively applied for the numerical simulation of the electromagnetic-mechanical coupling induced in the last design configuration of NET first wall during a plasma disruption event. A picture of the impact of the electromagnetic forces on the structural behaviour of the outboard DN first wall is presented an an improvement of the FW structural section is proposed. In any case, additional investigations will be performed during the long process of structural behaviour optimization of the first wall reactor components

Damage of first wall materials in fusion reactors under nonstationary thermal effects

International Nuclear Information System (INIS)

Maslaev, S.A.; Platonov, Yu.M.; Pimenov, V.N.

1991-01-01

The temperature distribution in the first wall of a fusion reactor was calculated for nonstationary thermal effects of the type of plasma destruction or the flow of 'running electrons' taking into account the melting of the surface layer of the material. The thickness of the resultant damaged layer in which thermal stresses were higher than the tensile strength of the material is estimated. The results were obtained for corrosion-resisting steel, aluminium and vanadium. Flowing down of the molten layer of the material of the first wall is calculated. (author)

First wall and blanket design for the STARFIRE commercial tokamak power reactor

International Nuclear Information System (INIS)

Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

1979-01-01

The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

Materials issues in the design of the ITER first wall, blanket, and divertor

International Nuclear Information System (INIS)

Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

1992-01-01

During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

Numerical analysis of heat transfer in the first wall of CFETR WCSB blanket

Energy Technology Data Exchange (ETDEWEB)

Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Deng, Weiping; Ge, Zhihao; Li, Yuanjie

2016-04-15

Highlights: • Detailed numerical analysis of heat transfer in a water-cooling first wall was carried out based on the conceptual design of CFETR WCSB blanket. • Investigation of the influences of buoyancy effect and surface roughness on heat transfer in the water-cooling first wall was presented. • Analysis of the effect of the front wall thickness on temperature was carried out for the water-cooling first wall design. • Simulation results of two 1D CFD methods were evaluated by the 3D CFD data. - Abstract: China Fusion Engineering Test Reactor (CFETR), the first fusion reactor experiment project planned in China, is now being investigated in detail. Recently, a conceptual structural design of the Water-Cooled-Solid-Breeder (WCSB) blanket was proposed as one of the breeding blanket candidates for CFETR. In this research, based on the present design of the CFETR WCSB blanket, the heat transfer performance in the first wall (FW) under the pressurized water cooling condition was analyzed. The 3D computational fluid dynamics (CFD) results show that the maximal temperature of the FW will not exceed the limited temperature under normal or even higher heat flux condition. In addition, the effect of buoyancy on heat transfer is negligible under both conditions. The influence of roughness becomes increasingly important when the roughness height lies in the fully turbulent regime. The maximal temperature increases approximately linearly as the thickness of the front wall increases. It is also found that the heat flux and the local heat transfer coefficient are extremely non-uniform in the circumferential direction. Two 1D CFD methods are also evaluated by 3D CFD data, with the conclusion that both 1D results have some differences with the 3D data. The improved 1D method is more accurate than the former one. However, we ascertain that 1D methods should be used with caution for the water-cooling FW design.

Condensation of ablated first-wall materials in the cascade inertial confinement fusion reactor

International Nuclear Information System (INIS)

Ladd, A.J.C.

1985-01-01

This report concerns problems involved in recondensing first-wall materials vaporized by x rays and pellet debris in the Cascade inertial confinement fusion reactor. It examines three proposed first-wall materials, beryllium oxide (BeO), silicon carbide (SiO), and pyrolytic graphite (C), paying particular attention to the chemical equilibrium and kinetics of the vaporized gases. The major results of this study are as follows. Ceramic materials composed of diatomic molecules, such as BeO and SiC, exist as highly dissociated species after vaporization. The low gas density precludes significant recombination during times of interest (i.e., less than 0.1 s). The dissociated species (Be, O, Si, and C) are, except for carbon, quite volatile and are thermodynamically stable as a vapor under the high temperature and low density found in Cascade. These materials are thus unsuitable as first-wall materials. This difficulty is avoided with pyrolytic graphite. Since the condensation coefficient of monatomic carbon vapor (approx. 0.5) is greater than that of the polyatomic vapor (<0.1), recondensation is assisted by the expected high degree of dissociation. The proposed 10-layer granular carbon bed is sufficient to condense all the carbon vapor before it penetrates to the BeO layer below. The effective condensation coefficient of the porous bed is about 50% greater than that of a smooth wall. An estimate of the mass flux leaving the chamber results in a condensation time for a carbon first wall of about 30 to 50 ms. An experiment to investigate condensation in a Cascade-like chamber is proposed

First-wall-coating candidates for ICF reactor chambers using dry-wall protection only

International Nuclear Information System (INIS)

Sink, D.A.

1983-01-01

Twenty pure metals were considered as potential candidates for first-wall coatings of ICF reactor chambers. Seven were found to merit further consideration based on the results of computer-code calculations of figures-of-merit. The seven are rhenium, iridium, molybdenum, chromium, tungsten, tantalum, and niobium (listed in order of decreasing values of figures-of-merit). The calculations are based on mechanical, thermal, and vacuum vaporization engineering constraints. A number of alloys of these seven metals are suggested as additional candidates

A comparison of hydrogen vs. helium glow discharge effects on fusion device first-wall conditioning

International Nuclear Information System (INIS)

Dylla, H.F.

1989-09-01

Hydrogen- and deuterium-fueled glow discharges are used for the initial conditioning of magnetic fusion device vacuum vessels following evacuation from atmospheric pressure. Hydrogenic glow discharge conditioning (GDC) significantly reduces the near-surface concentration of simple adsorbates, such as H 2 O, CO, and CH 4 , and lowers ion-induced desorption coefficients by typically three orders of magnitude. The time evolution of the residual gas production observed during hydrogen-glow discharge conditioning of the carbon first-wall structure of the TFTR device is similar to the time evolution observed during hydrogen GDC of the initial first-wall configuration in TFTR, which was primarily stainless steel. Recently, helium GDC has been investigated for several wall-conditioning tasks on a number of tokamaks including TFTR. Helium GDC shows negligible impurity removal with stainless steel walls. For impurity conditioning with carbon walls, helium GDC shows significant desorption of H 2 O, CO, and CO 2 ; however, the total desorption yield is limited to the monolayer range. In addition, helium GDC can be used to displace hydrogen isotopes from the near-surface region of carbon first-walls in order to lower hydrogenic retention and recycling. 38 refs., 6 figs

Results of strategic calculations for optimizing the first wall life in a tokamak fusion reactor

International Nuclear Information System (INIS)

Daenner, W.

1981-01-01

The development of the FWLTB computer program has reached a stage where prediction of the first wall lifetime is possible. Because of the large number of free parameters strategic calculations were found to be the most appropriate way to arrive at load design conditions which allow optimum life expectancy. In this paper a revised set of life criteria is presented this being followed by the results of parameter studies in which single parameters were varied while the remaining ones were kept fixed at a reference value. These results are used as a guide during the subsequent strategic calculations. In a first strategy we aimed at finding the maximum lifetime for the case that the reactor is operated at a neutron wall loading of 10 MW/m 2 . We found that operation over a period of more than one year is possible if the first wall is designed in a very tiny geometry and cooled by a low-pressure coolant. In a second strategy the aim was to find the design conditions for the case that the first wall is cooled by a high-pressure coolant. It is shown that liquid-lithium cooling is manageable up to high wall loadings, but the lifetime is restricted to about 6 MWa/m 2 . Helium cooling allows a higher lifetime, but the design conditions are such that only modest wall loadings can be permitted. (orig.)

First wall material damage induced by fusion-fission neutron environment

Energy Technology Data Exchange (ETDEWEB)

Khripunov, Vladimir, E-mail: Khripunov_VI@nrcki.ru

2016-11-01

Highlights: • The highest damage and gas production rates are experienced within the first wall materials of a hybrid fusion-fission system. • About ∼2 times higher dpa and 4–5 higher He appm are expected compared to the values distinctive for a pure fusion system at the same DT-neutron wall loading. • The specific nuclear heating may be increased by a factor of ∼8–9 due to fusion and fission neutrons radiation capture in metal components of the first wall. - Abstract: Neutronic performance and inventory analyses were conducted to quantify the damage and gas production rates in candidate materials when used in a fusion-fission hybrid system first wall (FW). The structural materials considered are austenitic SS, Cu-alloy and V- alloys. Plasma facing materials included Be, and CFC composite and W. It is shown that the highest damage rates and gas particles production in materials are experienced within the FW region of a hybrid similar to a pure fusion system. They are greatly influenced by a combined neutron energy spectrum formed by the two-component fusion-fission neutron source in front of the FW and in a subcritical fission blanket behind. These characteristics are non-linear functions of the fission neutron source intensity. Atomic displacement damage production rate in the FW materials of a subcritical system (at the safe subcriticality limit of ∼0.95 and the neutron multiplication factor of ∼20) is almost ∼2 times higher compared to the values distinctive for a pure fusion system at the same 14 MeV neutron FW loading. Both hydrogen (H) and helium (He) gas production rates are practically on the same level except of about ∼4–5 times higher He-production in austenitic and reduced activation ferritic martensitic steels. A proper simulation of the damage environment in hybrid systems is required to evaluate the expected material performance and the structural component residence times.

Thermal and radiation loads on the first wall and divertor plates in the KTM tokamak

International Nuclear Information System (INIS)

Azizov, Eh.A.; Buzhinskij, O.I.; Gladush, G.G.; Darmagraj, V.V.; Priyampol'skij, I.R.; Dvorkin, N.Ya.; Lejkin, I.N.; Tazhibaeva, I.L.; Shestakov, V.P.

2001-01-01

The constructing of the KTM tokamak is intended for wide scale studies of behavior both inner-chamber element materials and structures (first wall, limiters, divertor, hf-antennas, etc.) under conditions approaching to the ITER-FEAT and a future thermonuclear reactors. The KTM tokamak is designed for maintain of interaction conditions of plasma-wall, plasma flows and divertor field, stimulating conditions of ITER-FEAT; and for examination of a future tokamaks' materials. In the work the thermal loads on the first wall, divertor plates are presented

First wall of thermonuclear device

International Nuclear Information System (INIS)

Miki, Nobuharu.

1992-01-01

In a first wall of a thermonuclear device, armour tiles are metallurgically bonded to a support substrate only for the narrow area of the central portion thereof, while bonded by metallurgical bonding with cooling tubes of low mechanical toughness, separated from each other in other regions. Since the bonding area with the support substrate of great mechanical rigidity is limited to the narrow region at the central portion of the armour tiles, cracking are scarcely caused at the end portion of the bonding surface. In other regions, since cooling tubes of low mechanical rigidity are bonded metallurgically, they can be sufficiently withstand to high thermal load. That is, even if the armour tiles are deformed while undergoing thermal load from plasmas, since the cooling tubes absorb it, there is no worry of damaging the metallurgically bonded face. Since the cooling tubes are bonded directly to the armour tiles, they absorb the heat of the armour tiles efficiently. (N.H.)

Simulation of fusion first-wall environment in a fission reactor

International Nuclear Information System (INIS)

Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

1982-01-01

A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

Development of Joining Technologies for the ITER Blanket First Wall

Energy Technology Data Exchange (ETDEWEB)

Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of)

2011-01-15

The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

Development of Joining Technologies for the ITER Blanket First Wall

International Nuclear Information System (INIS)

Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon

2011-01-01

The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

Preliminary investigation on welding and cutting methods for first wall support leg in ITER blanket module

International Nuclear Information System (INIS)

Mohri, Kensuke; Suzuki, Satoshi; Enoeda, Mikio; Kakudate, Satoshi; Shibanuma, Kiyoshi; Akiba, Masato

2006-08-01

Concept of a module type of blanket has been applied to ITER shield blanket, of which size is typically 1mW x 1mH x 0.4mB with the weight of 4 ton, in order to enhance its maintainability and fabricability. Each shield blanket module consists of a shield block and four first walls which are separable from the shield block for the purpose of reduction of an electro-magnetic force in disruption events, radio-active waste reduction in the maintenance work and cost reduction in fabrication process. A first wall support leg, a part of the first wall component located between the first wall and the shield block, is required not only to be connected metallurgically to the shield block in order to withstand the electro-magnetic force and coolant pressure, but also to be able to replace the first wall more than 2 times in the hot cell during the life time of the reactor. Therefore, the consistent structure where remote handling equipment can be access to the joint and carry out the welding/cutting works perfectly to replace the first wall in the hot cell is required in the shield blanket design. This study shows an investigation of the blanket module no.10 design with a new type of the first wall support leg structure based on Disc-Cutter technology, which had been developed for the main pipe cutting in the maintenance phase and was selected out of a number of candidate methods, taking its large advantages into account, such as 1) a post-treatment can be eliminated in the hot cell because of no making material chips and of no need of lubricant, 2) the cut surface can be rewelded without any machining. And also, a design for the small type of Disc-Cutter applied to the new blanket module no.10 has been investigated. In conclusion, not only the good performance of Disc-Cutter technology applied to the updated blanket module, but also consistent structure of the simplified shield blanket module including the first wall support leg in order to satisfy the requirements in the

Summary of beryllium qualification activity for ITER first-wall applications

International Nuclear Information System (INIS)

Barabash, V; Eaton, R; Hirai, T; Kupriyanov, I; Nikolaev, G; Wang Zhanhong; Liu Xiang; Roedig, M; Linke, J

2011-01-01

Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

Summary of beryllium qualification activity for ITER first-wall applications

Science.gov (United States)

Barabash, V.; Eaton, R.; Hirai, T.; Kupriyanov, I.; Nikolaev, G.; Wang, Zhanhong; Liu, Xiang; Roedig, M.; Linke, J.

2011-12-01

Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

Deuterium behavior in first-wall materials for nuclear fusion

International Nuclear Information System (INIS)

Bernard, E.

2012-01-01

Plasma-wall interactions play an important part while choosing materials for the first wall in future fusion reactors. Moreover, the use of tritium as a fuel will impose safety limits regarding the total amount present in the tokamak. Previous analyses of first-wall samples exposed to fusion plasma highlighted an in-bulk migration of deuterium (as an analog to tritium) in carbon materials. Despite its limited value, this retention is problematic: contrary to co-deposited layers, it seems very unlikely to recover easily the deuterium retained in such a way. Because of the difficult access to in situ samples, most published studies on the subject were carried out using post-mortem sample analysis. In order to access to the dynamic of the phenomenon and come apart potential element redistribution during storage, we set up a bench intended for simultaneous low-energy ion implantation, reproducing the deuterium interaction with first-wall materials, and high-energy micro beam analysis. Nuclear reaction analysis performed at the micrometric scale (μNRA) allows to characterize deuterium repartition profiles in situ. This analysis technique was confirmed to be non-perturbative of the mechanisms studied. We observed on the experimental data set that the material surface (0-1 μm) display a high and nearly constant deuterium content, with a uniform distribution. On the contrary, in-bulk deuterium (1-11 μm) localizes in preferential trapping sites related to the material microstructure. In-bulk deuterium inventory seems to increase with the incident fluence, in spite of the wide data scattering attributed to the structure variation of studied areas. Deuterium saturation at the surface as well as in-depth migration are instantaneous; in-vacuum storage leads to a small deuterium global desorption. Observations made via μNRA were coupled with results from other characterization techniques. X-ray μtomography allowed to identify porosities as the preferential trapping sites

Development of vanadium base alloys for fusion first-wall/blanket applications

International Nuclear Information System (INIS)

Smith, D.L.; Chung, H.M.; Loomis, B.A.; Matsui, H.; Votinov, S.; VanWitzenburg, W.

1994-01-01

Vanadium alloys have been identified as a leading candidate material for fusion first-wall/blanket applications. Certain vanadium alloys exhibit favorable safety and environmental characteristics, good fabricability, high temperature and heat load capability, good compatibility with liquid metals and resistance to irradiation damage effects. The current focus is on vanadium alloys with (3-5)% Cr and (3-5)% Ti with a V-4Cr-4Ti alloy as the leading candidate. Preliminary results indicate that the crack-growth rates of certain alloys are not highly sensitive to irradiation. Results from the Dynamic Helium Charging Experiment (DHCE) which simulates fusion relevant helium/dpa ratios are similar to results from neutron irradiated material. This paper presents an overview of the recent results on the development of vanadium alloys for fusion first wall/blanket applications

The JET ITER-like wall experiment: First results and lessons for ITER

Energy Technology Data Exchange (ETDEWEB)

Horton, Lorne, E-mail: Lorne.Horton@jet.efda.org [EFDA-CSU Culham, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium)

2013-10-15

Highlights: ► JET has recently completed the installation of an ITER-like wall. ► Important operational aspects have changed with the new wall. ► Initial experiments have confirmed the expected low fuel retention. ► Disruption dynamics have change dramatically. ► Development of wall-compatible, ITER-relevant regimes of operation has begun. -- Abstract: The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.

High-flux first-wall design for a small reversed-field pinch reactor

International Nuclear Information System (INIS)

Cort, G.E.; Graham, A.L.; Christensen, K.E.

1982-01-01

To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high-technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time-averaged heat flux of 4.5 MW/m 2 with a conservatively estimated transient peak approximately twice the average value. We present the design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities. We combined a thermal analysis with a structural fatigue analysis to design the heat transfer module to last 10 6 cycles or one year at 80% duty for a 26-s power cycle. This fatigue life is compatible with a radiation damage life of 14 MW/yr/m 2

The first installation of the WindWall in the Netherlands; Eerste WindWall in Nederland geplaatst

Energy Technology Data Exchange (ETDEWEB)

Ten Bolscher, G.H.; Vander Heide, H. [DWA Installatie- en energieadvies, Bodegraven (Netherlands)

2004-02-01

This article is the first in a series of four on the experiment with the WindWall, a wind turbine on the roof of a school building in Zwolle, Netherlands. The experiment started September 5, 2003. [Dutch] Door diverse marktpartijen worden momenteel kleinere, voor de gebouwde omgeving geschikte windturbines ontwikkeld, die de negatieve eigenschappen van grote windturbines (waarschijnlijk) niet hebben. Hierbij gaat het om eigenschappen als geluidsbelasting, beschaduwing, zichtbare aanwezigheid en visuele vervuiling van het vrije landschap. Op 11 juli 2003 is de eerste WindWall, een 'liggende' windturbine, geplaatst op het dak van het Deltion college in Zwolle in het kader van een praktijkexperiment, dat gesubsidieerd wordt door de Provincie Overijssel. Op 5 september 2003 is het systeern officieel in gebruik genomen.

Lifetime analysis for fusion reactor first walls and divertor plates

International Nuclear Information System (INIS)

Horie, T.; Tsujimura, S.; Minato, A.; Tone, T.

1987-01-01

Lifetime analysis of fusion reactor first walls and divertor plates is performed by (1) a one-dimensional analytical plate model, and (2) a two-dimensional elastic-plastic finite element method. Life-limiting mechanisms and the limits of applicability for these analysis methods are examined. Structural design criteria are also discussed. (orig.)

Technical issues and solutions on ITER first wall beryllium application. Industrial viewpoint

International Nuclear Information System (INIS)

Iwadachi, T.; Uda, M.; Ito, M.; Miyakawa, M.; Ibuki, M.

2004-01-01

Beryllium is selected as reference armor material of ITER primary first wall and is joined to the copper alloy heat sink such as CuCrZr or Dispersion Strengthened Copper (DSCu) Various joining technologies have been successfully developed and the manufacturing possibilities of large size first wall panels with beryllium armor has been demonstrated. Based on such results, further technical improvement is needed to reduce manufacturing cost and ensure the reliability of joining in actual size first wall. The technical issues to optimize the fabrication process of beryllium attachment were shown in this paper from an industrial point of view. Determination of the optimum size and the surface qualities of beryllium tiles are important issues in term of the material specification to ensure joining reliability and to reduce cost. The consolidation method and the finish machining methods of beryllium tiles are also critical in terms of material cost. These items should be determined by paying concern to the accommodation of the joining methods. The selections of slitting methods for attached beryllium have a great influence on fabrication cost. In the actual fabrication of beryllium attachment, safety provisions for exposure to beryllium in working environment and the recycling of the waste from the fabrication processes will be concerned sufficiently. (author)

Study on flow instability for feasibility of a thin liquid film first wall

Energy Technology Data Exchange (ETDEWEB)

Okino, Fumito, E-mail: fumito.okino@iae.kyoto-u.ac.jp [Kyoto University Graduate School of Energy Science, Gokasho Uji, Kyoto (Japan); Kasada, Ryuta; Konishi, Satoshi [Kyoto University Institute of Advanced Energy, Gokasho Uji, Kyoto (Japan)

2014-10-15

Highlights: • We propose a probability of an instability wave growth on a liquid metal first wall. • Evaporated gas by the high energy flux is predicted to agitate this instability wave. • Liquid Pb-17Li with a velocity 10 m/s, the ambient gas must be below 6.2 × 10{sup 3} Pa. • This pressure corresponds to 1600 K and it is attainable under a fusion energy flux. • This probability is not yet verified so the full verifications are to be performed. - Abstract: This study proposes a probability of the evaporated gas that agitates a growing instability wave in a thin liquid film first wall. The liquid first wall was considered to be in vacuum and the effect of the ambient gas was neglected but the evaporated gas by the high energy fluxes is a probable cause of unstable wave agitation. The criterion is approximately expressed by the density ratio (Q{sub 2}) and the Weber number (We) as Q{sub 2} × We{sup 0.5} ≈ 5 × 10{sup −4}. Performed indirect experimental supported this criterion. For a case study of liquid Pb-17Li film with a velocity of 10 m/s, the evaporated gas pressure must be below 6.2 × 10{sup 3} Pa to maintain stable conditions. By recent study, this pressure is generated at 1600 K temperature and it is believed to be attainable by the energy fluxes on the first wall. This result is so far not confirmed so the full verification by experimental is to be performed.

Measurement and modification of first-wall surface composition in the Oak Ridge Tokamak (ORMAK)

International Nuclear Information System (INIS)

Clausing, R.E.; Emerson, L.C.; Heatherly, L.; Colchin, R.J.; Twichell, J.C.

1975-01-01

Impurities coming into the plasma from the walls of present-day toroidal plasma confinement devices modify plasma behavior substantially. Small fractions of high-Z ions in the plasma greatly decrease plasma temperatures and increase plasma energy losses. Impurities from the ''first-wall'' in ORMAK were studied. Auger electron spectroscopy, soft x-ray appearance potential spectroscopy, and other surface sensitive techniques were used to characterize the surface composition of the first wall and to develop methods to remove carbon and oxygen. Oxygen glow discharge cleaning has been shown, in the laboratory, to be an effective way of removing carbon from gold films (simulated ORMAK linear material) and the use of oxygen discharge cleaning in ORMAK has resulted in a decrease in plasma contamination, a 50 percent increase in plasma current and an accompanying increase in plasma temperature. In spite of these improvements the walls of ORMAK are far from clean. Substantial amounts of carbon, oxygen, iron and other elements remain. (auth)

Operation experiences of the JT-60 first walls during high-power additional heating experiments

International Nuclear Information System (INIS)

Takatsu, H.; Ando, T.; Yamamoto, M.; Arai, T.; Kodama, K.; Suzuki, M.; Shimizu, M.

1989-01-01

JT-60 started its operation in May 1985 with TiC-coated molybdenum or Inconel 625 first walls. They provided very clean surfaces as well as superior plasma characteristics during Joule heating discharges. Though 20 μm-thick TiC coatings showed good adhesion characteristics, melting of the TiC coating and also the molybdenum or Inconel 625 substrate was observed at some specific spots, and an influx of heavy metals to the main plasma was inevitable during discharges. Initial results of the additional heating experiments showed degrading effects of locally melted TiC-coated molybdenum or Inconel 625 on plasma operation. Therefore, about a half of the TiC-coated first walls were removed and new graphite first walls were installed during the venting period from April to May 1987. The start-up of the discharge conditioning after installation of a significant number of graphite tiles was very rapid. Flexibility in plasma operation was increased, and JT-60 extended the operation region beyond its original specifications. The graphite first walls of the main chamber performed admirably and maintained their integrity under the conditions of plasma current and additional heating power up to 3.2 MA and 30 MW, respectively. On the other hand, the number of damaged divertor plates was much larger than that expected. The reason of unexpected failure is now under examination. (orig.)

Plasma Chamber and First Wall of the Ignitor Experiment^*

Science.gov (United States)

Cucchiaro, A.; Coppi, B.; Bianchi, A.; Lucca, F.

2005-10-01

The new designs of the Plasma Chamber (PC) and of the First Wall (FW) system are based on updated scenarios for vertical plasma disruption (VDE) as well as estimates for the maximum thermal wall loadings at ignition. The PC wall thickness has been optimized to reduce the deformation during the worst disruption event without sacrificing the dimensions of the plasma column. A non linear dynamic analysis of the PC has been performed on a 360^o model of it, taking into account possible toroidal asymmetries of the halo current. Radial EM loads obtained by scaling JET measurements have been also considered. The low-cycle fatigue analysis confirms that the PC is able to meet a lifetime of few thousand cycles for the most extreme combinations of magnetic fields and plasma currents. The FW, made of Molybdenum (TZM) tiles covering the entire inner surface of the PC, has been designed to withstand thermal and EM loads, both under normal operating conditions and in case of disruption. Detailed elasto-plastic structural analyses of the most (EM) loaded tile-carriers show that these are compatible with the adopted fabrication requirements. ^*Sponsored in part by ENEA of Italy and by the U.S. DOE.

First wall and shield components manufacturing by hot isostatic pressing

International Nuclear Information System (INIS)

Lind, Anders; Tegman, R.

1994-01-01

At a meeting in Garching in June 1994 Hot Isostatic Pressing (HIP) was presented as a possible route to manufacture ITER first wall and shield components. The main advantages of the HIP concept include excellent and uniform mechanical properties of the produced materials and joints, high reliability and robustness of the HIP process, double containment of coolant, good flexibility concerning general design as well as size and location for inner cooling tubes, low cost and short delivery times, and a good near net shape capability for components in size up to 15 tons. To assess the applicability of HIP for the manufacturing of ITER first wall and shield components, it was agreed * to choose possible production parameters based in the present know-how, * to produce a compound mock-up in one shot from available solid steel/powder copper/steel tubes to demonstrate the joinability of the materials, * to examine the produced mock-up/materials by multi array ultrasonic testing, limited mechanical testing, metallography, scanning electron microscopy and energy dispersive spectroscopy, and * to compile data on Type 316L steels produced by HIP. Preliminary results and the mock-up were presented at a meeting in Garching in mid July 1994. This study clearly shows the excellent joinability of a copper alloy (Cu-0.5%Zr) and stainless steels (Type 304, 316 L) by HIP at temperatures close to the melting temperature of copper, with only limited influence on the microstructures, which makes it possible to HIP the first wall and shield structure in one step. Excellent mechanical properties of the compound are obtained with the copper alloy and not the joint being the weakest part. 7 refs, 21 figs, 1 tab

Electron beam disruption simulation of first wall material

International Nuclear Information System (INIS)

Quataert, D.; Brossa, F.; Moretto, P.; Rigon, G.

1984-01-01

The destructive effect of plasma disruptions on first wall material and limiters has been predicted and models have been made to study their behaviour under intensive pulsed energy deposition. The results presented here give a full description of qualitative and semi-quantitative results obtained for several materials (Mo, stainless steel, Cu, Al, Inconel, etc.) under various experimental conditions. Examples are given of specific defects such as: evaporation, melting, void and crack formation and recrystallization of the underlying material. Methods for the evaluation of deposited energy and beam dimensions are also presented. (author)

Summary report for IAEA CRP on lifetime prediction for the first wall of a fusion machine (JAERI contribution)

International Nuclear Information System (INIS)

Suzuki, Satoshi; Araki, Masanori; Akiba, Masato

1993-03-01

IAEA Coordinated Research Program (CRP) on 'Lifetime Prediction for the First Wall of a Fusion Machine' was started in 1989. Five participants, Joint Research Centre (JRC-Ispra), The NET team, Kernforschungszentrum Karlsruhe (KfK), Russian Research Center and Japan Atomic Energy Research Institute, contributed in this activity. The purpose of the CRP is to evaluate the thermal fatigue behavior of the first wall of a next generation fusion machine by means of numerical methods and also to contribute the design activities for ITER (International Thermonuclear Experimental Reactor). Thermal fatigue experiments of a first wall mock-up which were carried out in JRC-Ispra were selected as a first benchmark exercise model. All participants performed finite element analyses with various analytical codes to predict the lifetime of the simulated first wall. The first benchmark exercise has successfully been finished in 1992. This report summarizes a JAERI's contribution for this first benchmark exercise. (author)

Refractory oxides for fusion reactor first walls, the effects of the reducing environment

International Nuclear Information System (INIS)

Hoffman, J.G.

1979-01-01

Of the several applications for refractory oxides in fusion reactor systems, the most demanding is that for the first wall. Some components in proximity of the first wall (possibly waveguides or flux breakers) will also be subjected to similar environments. Many parameters affect the ultimate usability of a particular material for reactor applications: electrical resistivity and dielectric breakdown if applicable, thermal conductivity, mechanical properties, and stability with respect to neutral molecular or atomic, or ionized fuel gases. All these properties can be affected by the radiation environment present in an operating power reactor. Temperatures up to 2000K may be expected for radiatively cooled first wall liners in some proposed designs although surface temperatures are appreciably lower (approximately 1000K) in other applications. The exact nature of the chemical environment is not defined even for the most well developed design concepts, but possible environments may be hypothesized; ambient neutral molecular and atomic species, bombardment by high energy charge exchange neutral atoms, direct ionic bombardment from stray ions, and plasma dumps from failure of the confinement system. Preliminary work has begun to more adequately define the extent of the problem and suggest approaches to engineering solutions

Diagnostic techniques for measuring temperature transients and stress transients in the first wall of an ICF reactor

International Nuclear Information System (INIS)

Melamed, N.T.; Taylor, L.H.

1983-01-01

The primary challenge in the design of an Inertial Confinement Fusion (ICF) power reactor is to make the first wall survive the frequent explosions of the pellets. Westinghouse has proposed a dry wall design consisting of steel tubes coated with tantalum. This report describes the design of a test chamber and two diagnostic procedures for experimentally determining the reliability of the Westinghouse design. The test chamber simulates the x-ray and ion pulse irradiation of the wall due to a pellet explosion. The diagnostics consist of remote temperature sensing and surface deformation measurements. The chamber and diagnostics can also be used to test other first-wall designs

Heat transfer modelling of first walls subject to plasma disruption

International Nuclear Information System (INIS)

Fillo, J.A.; Makowitz, H.

1981-01-01

A brief description of the plasma disruption problem and potential thermal consequences to the first wall is given. Thermal models reviewed include: a) melting of a solid with melt layer in place; b) melting of a solid with complete removal of melt (ablation); c) melting/vaporization of a solid; and d) vaporization of a solid but no phase change affecting the temperature profile

First-wall and blanket engineering development for magnetic-fusion reactors

International Nuclear Information System (INIS)

Baker, C.; Herman, H.; Maroni, V.; Turner, L.; Clemmer, R.; Finn, P.; Johnson, C.; Abdou, M.

1981-01-01

A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories

Engineering design and performances of the IGNITOR first wall

International Nuclear Information System (INIS)

Bonizzoni, G.

1989-01-01

Extensive work was carried out to define the working conditions and the reference design of the first wall for the IGNITOR machine: graphite covered modular elements attached to the vacuum vessel by a locking key for remote handling are proposed. The work includes a transient thermostructural analysis of the graphite tiles to evaluate temperatures and thermal stresses in normal and fault conditions. A full scale prototype of the element was manufactured. (author). 7 figs.; 1 tab

Radiation loads on the ITER first wall during massive gas injection

Energy Technology Data Exchange (ETDEWEB)

Landman, I., E-mail: igor.landman@kit.edu [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Bazylev, B. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Saibene, G. [Fusion for Energy Joint Undertaking, Josep Pla no. 2 – Torres Diagonal Litoral Edificio B3 7/03, Barselona 08019 (Spain); Pestchanyi, S. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Putvinski, S.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: • The massive gas injection (neon) is simulated with the two-dimensional tokamak code TOKES assuming the toroidal symmetry. • The neon injection, assimilation and transport of impurities through the entire plasma volume are modelled. • The output of TOKES is used by the melt motion code MEMOS to assess beryllium wall temperature and the regime with melting. • Complete plasma cooling occurs in minimum time of 5.7 ms with avoiding Be melting at any point on the first wall. -- Abstract: Unmitigated disruptions in ITER can produce strong localized surface damage on the first wall (FW). Massive gas injection (MGI) systems are being designed to dissipate a large fraction of the plasma stored energy at the disruption thermal quench (TQ) and hence reduce the consequences for FW components. The stored energies can be high enough, however, for there to be potential for the photon flash at the MGI TQ to drive local melting of beryllium FW components. To estimate the poloidal distribution of FW surface temperatures, the MGI process is being simulated using the 2D code TOKES, assuming toroidal symmetry. High pressure neon injection, assimilation and transport of injected impurities through the entire plasma volume are modelled. The output of these simulations is used by the melt motion code MEMOS to assess the resulting maximum surface temperature and the regimes with melting on the FW surface.

Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

International Nuclear Information System (INIS)

Conlin, J.A.

1985-08-01

A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

First wall for thermonuclear device

International Nuclear Information System (INIS)

Shibuya, Yoji.

1988-01-01

Purpose: To reduce the thermal stresses resulted to tiles and suppress the temperature rise for mounting jigs in first walls for a thermonuclear device. Constitution: A support mounting rod as a tile mounting and fixing jig and a fixing support connected therewith are disposed to the inside of an armour tile composed of high melting material and, further, a spring is disposed between the lower portion of the tile and the base plate. The armour tile can easily be fixed to the base plate by means of the resilient member by rotating the support member and abutting the support member against the support member abutting portion of the base plate. Further, since the contact and fixing surface of the armour tile and the fixing jig is situated below the tile inside the cooled base plate, the temperature rise can be suppressed as compared with the usual case. Since screw or like other clamping portion is not used for fixing the tile, heat resistant ceramics can be used with no restriction only to metal members, to thereby moderate the restriction in view of the temperature. (Kamimura, M.)

Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components

Energy Technology Data Exchange (ETDEWEB)

Barrett, Thomas R., E-mail: tom.barrett@ukaea.uk [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Ellwood, G.; Pérez, G.; Kovari, M.; Fursdon, M.; Domptail, F.; Kirk, S.; McIntosh, S.C.; Roberts, S.; Zheng, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Boccaccini, L.V. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); You, J.-H. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Bachmann, C. [EUROfusion, PPPT, Boltzmann Str. 2, 85748 Garching (Germany); Reiser, J.; Rieth, M. [KIT, IAM, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Visca, E.; Mazzone, G. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Arbeiter, F. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Domalapally, P.K. [Research Center Rez, Hlavní 130, 250 68 Husinec – Řež (Czech Republic)

2016-11-01

Highlights: • The engineering of the plasma facing components for DEMO is an extreme challenge. • PFC overall requirements, methods for assessment and designs status are described. • Viable divertor concepts for 10 MW/m{sup 2} surface heat flux appear to be within reach. • The first wall PFC concept will need to vary poloidally around the wall. • First wall coolant, structural material and PFC topology are open design choices. - Abstract: The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.

Response of ISSEC protected first walls to DT and DD plasma neutrons

International Nuclear Information System (INIS)

Avci, H.I.; Kulcinski, G.L.

1976-01-01

It has been demonstrated that the displacement damage and gas production rates can be reduced in CTR first walls by employing passive carbon shields. Reductions in displacement damage range from 3 to 5 for 12.5 cm shield thickness and from 7 to 14 in gas production rates with the same carbon thickness. The factors of reduction are 8 to 20 for the displacements and 17 to 80 for the gas production if a 25 cm shield is used. Depending on whether the isotopes causing the radioactivity are produced as a result of fast or thermal neutron activation, the first wall radioactivity can either go up or down with the increasing carbon shield thickness. It has been found that at shutdown radioactivity in 316 SS, Al, and Nb first walls is reduced with increasing carbon thickness while the activities in V and Ta are increased. Long term radioactivity displays the same trends in Al, 316 SS and Ta as short term radioactivity. However, the long term activity in Nb increases and that in V decreases with increasing shield thickness. It has also been found that systems operating on a D-D plasma cycle have higher displacement rates than respective D-T cycle systems. Gas production rates are slightly lower in D-D systems except for He production in 316 SS. This is due to the higher 59 Ni (n,α) cross sections for thermal neutrons

Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

International Nuclear Information System (INIS)

Merola, M.; Biggio, M.

1989-01-01

Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

The first installation of the WindWall in the Netherlands; Eerste WindWall in Nederland geplaatst

Energy Technology Data Exchange (ETDEWEB)

Ten Bolscher, G.H.; Vander Heide, H. [DWA Installatie- en energieadvies, Bodegraven (Netherlands)

2003-09-01

This article is the first in a series of four on the experiment with the WindWall, a wind turbine on the roof of a school building in Zwolle, Netherlands. The experiment started September 5, 2003. [Dutch] ledereen kent de grote windturbines die elektriciteit opwekken. Nadeel ervan is dat het draagvlak voor plaatsing op het land minder wordt, laat staan dat er mogelijkheden zijn voor toepassing in de gebouwde orngeving. Door diverse marktpartijen worden momenteel kleinere, voor de gebouwde omgeving geschikte windturbines ontwikkeld, die de negatieve eigenschappen van grote windturbines (waarschijnlijk) niet hebben. Hierbij gaat het om eigenschappen als geluidsbelasting, beschaduwing, zichtbare aanwezigheid en visuele vervuiling van het vrije landschap. Op 11 juli 2003 is de eerste WindWall, een 'liggende' windturbine, geplaatst op het dak van het Deltion college in Zwolle in het kader van een praktijkexperiment, dat gesubsidieerd wordt door de Provincie Overijssel. Op 5 september 2003 is het systeern officieel in gebruik genomen.

EU contribution to the procurement of the ITER blanket first wall

International Nuclear Information System (INIS)

Lorenzetto, Patrick; Banetta, Stefano; Bellin, Boris; Boireau, Bruno; Bucci, Philippe; Cicero, Tindaro; Conchon, Denis; Dellopoulos, Georges; Hardaker, Stephen; Marshall, Paul; Nogué, Patrice; Pérez, Marcos; Gutierrez, Leticia Ruiz; Samaniego, Fernando; Sherlock, Paul; Zacchia, Francesco

2016-01-01

Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

EU contribution to the procurement of the ITER blanket first wall

Energy Technology Data Exchange (ETDEWEB)

Lorenzetto, Patrick, E-mail: Patrick.Lorenzetto@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Banetta, Stefano; Bellin, Boris [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Boireau, Bruno [AREVA NP, Centre Technique, 71200 Le Creusot (France); Bucci, Philippe [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Cicero, Tindaro [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Conchon, Denis [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Dellopoulos, Georges [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Hardaker, Stephen [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Marshall, Paul [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Nogué, Patrice [AREVA NP, Centre Technique, 71200 Le Creusot (France); Pérez, Marcos [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Gutierrez, Leticia Ruiz [Iberdrola Ingeniería y Construcción S.A.U., Avenida Manoteras 20, 28050 Madrid (Spain); Samaniego, Fernando [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Sherlock, Paul [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Zacchia, Francesco [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain)

2016-11-01

Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

Fusion technology development: first wall/blanket system and component testing in existing nuclear facilities

International Nuclear Information System (INIS)

Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

1980-12-01

A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment employing an existing nuclear facility, the Engineering Test Reactor at the Idaho National Engineering Laboratory, is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of an existing test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module. The principal reaction considered involves 3 He in the annulus as follows: n + 3 He → p + t + 0.75 MeV. Bulk heating in the test module is accomplished by neutron thermalization, gamma heating, and absorption reactions involving 6 Li in the blanket breeding region. The concept can be extended to modified core configurations that will accommodate test modules of different sizes and types. It makes possible development testing of first wall/blanket systems and other fusion components on a scale and in ways not otherwise available until actual high-power fusion reactors are built

Neutral particle balance in GDT with fast titanium coating of the first wall

International Nuclear Information System (INIS)

Bagryansky, P.A.; Bender, E.D.; Ivanov, A.A.; Krahl, S.; Noack, K.; Karpushov, A.N.; Murakhtin, S.V.; Shikhovtsev, I.V.

1995-01-01

The GDT is an axisymmetric open trap with a high mirror ratio for confinement of a collisional plasma. The experimental program of the GDT was focused on the generation of plasma physics database necessary for a GDT-based neutron source. A distinct feature of both GDT and the GDT-based neutron source is that the Larmor radius of the fast sloshing ions is comparable to plasma radius. In this case, the sloshing ions can not be well shielded by the plasma halo from penetration of the neutral gas from periphery that results in high charge exchange losses. The plasma parameters are then very sensitive to gas pressure near the plasma boundary. To reduce the gas pressure to desured value during the beam heating, the authors have used arc-type evaporators developed at the Budker INP for fast titanium coating of the GDT first wall. If needed, the coating can be done a few seconds before each shot. They investigated the neutral particle balance in presence of NB-heating. The inverted magnetron gauges were used to study the temporal dependence of gas pressure inside the central cell. Pyroelectric bolometers were employed to measure the flux of charge exchange neutrals. Neutral particle balance has also been studied numerically by using a gas-transport code. The results of the investigations are the following: (1) sloshing ion lifetime was increased about 10 times compared to that without the coating of the first wall; and (2) wall recycling coefficient of the Ti-coated wall does not exceed 1 for 8 keV mean energy of the neutral hydrogen atoms striking the wall

Lifetime estimates of a fusion reactor first wall by linear damage summation and strain range partitioning methods

International Nuclear Information System (INIS)

Liu, K.C.; Grossbeck, M.L.

1979-01-01

A generalized model of a first wall made of 20% cold-worked steel was examined for neutron wall loadings ranging from 2 to 5 MW/m 2 . A spectrum of simplified on-off duty cycles was assumed with a 95% burn time. Independent evaluations of cyclic lifetimes were based on two methods: the method of linear damage summation currently being employed for use in ASME high-temperature design Code Case N-47 and that of strain range partitioning being studied for inclusion in the design code. An important point is that the latter method can incorporate a known decrease in ductility for materials subject to irradiation as a parameter, so low-cycle fatigue behavior can be estimated for irradiated material. Lifetimes predicted by the two methods agree reasonably well despite their diversity in concept. Lack of high-cycle fatigue data for the material tested at temperatures within the range of our interest precludes making conclusions on the accuracy of the predicted results, but such data are forthcoming. The analysis includes stress relaxation due to thermal and irradiation-induced creep. Reduced ductility values from irradiations that simulate the environment of the first wall of a fusion reactor were used to estimate the lifetime of the first wall under irradiation. These results indicate that 20% cold-worked type 316 stainless steel could be used as a first-wall material meeting a 8 to 10 MW-year/m 2 lifetime goal for a neutron wall loading of about 2 MW-year/m 2 and a maximum temperature of about 500 0 C

Tungsten as First Wall Material in Fusion Devices

International Nuclear Information System (INIS)

Kaufmann, M.

2006-01-01

In the PLT tokamak with a tungsten limiter strong cooling of the central plasma was observed. Since then mostly graphite has been used as limiter or target plate material. Only a few tokamaks (limiter: FTU, TEXTOR; divertor: Alcator C-Mod, ASDEX Upgrade) gained experience with high-Z-materials. With the observed strong co- deposition of tritium together with carbon in JET and as a result of design studies of fusion reactors, it became clear that in the long run tungsten is the favourite for the first-wall material. Tungsten as a plasma facing material requires intensive research in all areas, i.e. in plasma physics, plasma wall-interaction and material development. Tungsten as an impurity in the confined plasma reveals considerable differences to carbon. Strong radiation at high temperatures, in connection with mostly a pronounced inward drift forms a particular challenge. Turbulent transport plays a beneficial role in this regard. The inward drift is an additional problem in the pedestal region of H-mode plasmas in ITER-like configurations. The erosion by low energy hydrogen atoms is in contrast to carbon small. However, erosion by fast particles from heating measures and impurity ions, accelerated in the sheath potential, play an important role in the case of tungsten. Radiation by carbon in the plasma boundary reduces the load to the target plates. Neon or Argon as substitutes will increase the erosion of tungsten. So far experiments have demonstrated that in most scenarios the tungsten content in the central plasma can be kept sufficiently small. The material development is directed to the specific needs of existing or future devices. In ASDEX Upgrade, which will soon be a divertor experiment with a complete tungsten first-wall, graphite tiles are coated with tungsten layers. In ITER, the solid tungsten armour of the target plates has to be castellated because of its difference in thermal expansion compared to the cooling structure. In a reactor the technical

Development of laser-based technology for the routine first wall diagnostic on the tokamak EAST: LIBS and LIAS

Science.gov (United States)

Hu, Z.; Gierse, N.; Li, C.; Liu, P.; Zhao, D.; Sun, L.; Oelmann, J.; Nicolai, D.; Wu, D.; Wu, J.; Mao, H.; Ding, F.; Brezinsek, S.; Liang, Y.; Ding, H.; Luo, G.; Linsmeier, C.; EAST Team

2017-12-01

A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.

Effect of fusion burn cycle on first wall swelling

International Nuclear Information System (INIS)

Choi, Y.H.; Bement, A.L.; Russell, K.C.

1976-01-01

A mathematical simulation of first wall swelling has been performed for stainless steel under a hypothetical duty cycle of 50 sec burn, 50 sec cool. In most instances steady state nucleation conditions were not established during the burn cycle, thereby necessitating the use of transient nucleation theory. The effects of transmutation helium and of surface active impurities were modelled in an approximate way. Both kinds of impurity were found to give large increases in the void nucleation rate. Suggestions for refining and extending the calculations are also given

Relevance of NET first wall concept for DEMO DN

International Nuclear Information System (INIS)

Kiltie, J.S.

1987-01-01

Design studies for the Next European Torus (NET) have produced a design concept for the first wall. This concept features poloidal water cooling, double contained in a welded steel structure which is protected by radiatively cooled tiles. In this appendix the relevance of this concept to a DEMO is examined with particular emphasis given to the ability of the cooling tube arrangement to remove the heat. A suggested modification to the arrangement of coolant tubes is suggested so that the design can operate at the higher loadings of a DEMO. (author)

Progress in the design of mechanically attached, conductively cooled low-Z armour tiles for the NET integrated first wall

International Nuclear Information System (INIS)

Shaw, R.; Vieider, G.

1991-01-01

For the NET device complete or extensive coverage of the first wall with a low-Z armour is envisaged. This armour may comprise a general protection, ∝90% total first-wall surface, of low-temperature conductively cooled tiles, complemented by a local protection of radiatively cooled tiles in regions where near peak fluxes are incident. A low-temperature (∝1000deg C) carbon-based armour, cooled via conduction to the reference NET integrated first wall, has been developed using currently available materials. The armour comprises a small square tile fabricated in high-conductivity 3-D or random-fibre carbon fibre reinforced carbon composite attached to the steel first wall via a stainless-steel/refractory metal stud assembly. Attachment forces are maintained within acceptable limits, particularly during baking, through material selection and component geometry. To ensure effective heat transfer throughout the duty cycle an intermediate conductive layer of a highly compliant material is foreseen. The scope of the paper covers the design of the armour assembly for proof of principle testing with the NET first-wall test section, TS1, and reports the results of supporting thermomechanical analyses. (orig.)

Plasma surface engineering in first wall of tokamak

International Nuclear Information System (INIS)

Liu Xiang; Xu Zengyu; Zhang Fu; Zhang Nianman

2001-01-01

The boronization, siliconization and lithium coating of the inner wall of HL-1M are introduced, the hydrogen recycling and the influence to impurities controlled and core radiation energy loss are discussed. Experiments prove that these wall treatments are very useful for the plasma confinement, a 4 s reproducible long pulse discharge is obtained for siliconized wall, but the plasma pulse length only achieves 2.1 s and its reproducibility is very poor for boronized wall. Lithium coating is the best method of the wall treatments for lowering hydrogen recycling and decreasing the impurities level. For the applications of HL-2A and the future fusion device, a series of B, Ti, Si-doped graphite and B 4 C-C/C composites have been developed, some experimental results about chemical sputtering, tritium retention and recycling, as well as high heat loads are reviewed. Meanwhile, SiC, TiC and B 4 C coating, and B 4 C-C, SiC-C, B 4 C-Cu, Mo-Cu and W-Cu functionally graded materials are also introduced

Optimization of the first wall for the DEMO water cooled lithium lead blanket

Energy Technology Data Exchange (ETDEWEB)

Aubert, Julien, E-mail: julien.aubert@cea.fr [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Aiello, Giacomo [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Bachmann, Christian [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Di Maio, Pietro Alessandro [Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, Rosario [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy); Li Puma, Antonella; Morin, Alexandre [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Tincani, Amelia [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy)

2015-10-15

Highlights: • This paper presents the optimization of the first wall of the water cooled lithium lead DEMO blanket with pressurized water reactor condition and circular channels in order to find the best geometry that can allow the maximum heat flux considering design criteria since an estimate of the engineering limit of the first wall heat load capacity is an essential input for the decision to implement limiters in DEMO. • An optimization study was carried out for the flat first wall design of the DEMO Water-Cooled Lithium Lead considering thermal and mechanical constraint functions, assuming T{sub inlet}/T{sub outlet} equal to 285 °C/325 °C, based on geometric design parameters. • It became clear that through the optimization the advantages of a waved First Wall are diminished. • The analysis shows that the maximum heat load could achieve 2.53 MW m{sup −2}, but considering assumptions such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total first wall thickness ≤22 mm, heat flux is limited to 1.57 MW m{sup −2}. - Abstract: The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m{sup −2} with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet

Evaluation of copper alloys for fusion reactor divertor and first wall components

DEFF Research Database (Denmark)

Fabritsiev, S.A.; Zinkle, S.J.; Singh, B.N.

1996-01-01

This paper presents a critical analysis of the main factors of radiation damage limiting the possibility to use copper alloys in the ITER divertor and first wall structure. In copper alloys the most significant types of radiation damage in the proposed temperature-dose operation range are swellin...

The thermal response of the first wall of a fusion reactor blanket to plasma disruptions

International Nuclear Information System (INIS)

Klippel, H.Th.

1983-09-01

Major plasma disruptions in Tokamak power reactors are potentially dangerous because high thermal overloading of the first wall may occur, resulting in melting and evaporation. The present uncertainties of the disruption characteristics, in particular the space and time dependence of the energy deposition, lead to a wide variation in the prospective surface energy loads. The thermal response of a first wall of aluminium, stainless steel and of graphite subjected to disruption energy loads up to 1000 J cm -2 has been analysed including the effects of melting and surface evaporation, vapour recondensation, vapour shielding, and the moving of the surface boundary caused by the evaporation. A special calculation model has been developed for this purpose. The main results are the following: by values of local transient energy depositions over 1500 J cm -2 bare stainless steel walls are damaged severely. Further calculations are needed to estimate the endurance limit of several candidate first wall materials. Applications of coatings on surfaces need special attention. For the reference INTOR disruption (approx. 100 J cm -2 ) evaporation is not significant. The effect of vapour shielding on evaporation has been found to be significant. The effect on melting is less pronounced. In a complete analysis the stability and dynamic behaviour of the melted layer under electromagnetic forces should be included. Also a reliable set of plasma disruption characteristics should be gathered

First results from the 10Be marker experiment in JET with ITER-like wall

International Nuclear Information System (INIS)

Bergsåker, H.; Bykov, I.; Petersson, P.; Possnert, G.; Heinola, K.; Miettunen, J.; Groth, M.; Kurki-Suonio, T.; Widdowson, A.; Riccardo, V.; Nunes, I.; Stamp, M.; Brezinsek, S.; Borodin, D.; Kirschner, A.; Likonen, J.; Coad, J.P.; Schmid, K.; Krieger, K.

2014-01-01

When the ITER-like wall was installed in JET, one of the 218 Be inner wall guard limiter tiles had been enriched with 10 Be as a bulk isotopic marker. During the shutdown in 2012–2013, a set of tiles were sampled nondestructively to collect material for accelerator mass spectroscopy measurements of 10 Be concentration. The letter shows how the marker experiment was set up, presents first results and compares them to preliminary predictions of marker redistribution, made with the ASCOT numerical code. Finally an outline is shown of what experimental data are likely to become available later and the possibilities for comparison with modelling using the WallDYN, ERO and ASCOT codes are discussed. (letter)

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

A preliminary model for estimating the first wall lifetime of a fusion reactor

International Nuclear Information System (INIS)

Daenner, W.

1975-02-01

The estimation of the first wall lifetime is a necessary basis for predicting the availability of a fusion power plant. In order to do this, an analytical model was prepared and programmed for the computer which calculates the temperature and stress load of the first wall from the principal design parameters and quotes them against the relevant material properties. Neither the analytical model nor the information about the material performance is yet complete so that the answers obtained from the program are very preliminary. This situation is underlined by the results of sample calculations performed for the CTRD blanket module cell. The results obtained for vanadium and vanadium alloys show a strong dependence of the lifetime on the irradiation creep and the ductility of these materials. Completion of this model is envisaged as soon as the missing information becomes available. (orig.) [de

Temperature and displacement transients in inertial confinement fusion first-walls

International Nuclear Information System (INIS)

Hunter, T.O.; Kulcinski, G.L.

1977-01-01

A quasi-analytic general model is developed for determination of temperature response and displacement damage in materials exposed to bursts of thermonuclear radiations. Temperature response can be determined for any time or position. Materials are assessed, using the model, which might be employed for dry first walls, collectors, laser mirrors, or other exposed reactor components. The resulting magnitude and temporal distribution of temperature and displacement production show that effects on material micro-structure must be treated in a dynamic fashion

Erosion of the first wall of Tokamaks

International Nuclear Information System (INIS)

Guseva, M.I.; Ionova, E.S.; Martynenko, Yu.V.

1980-01-01

An estimate of the rate of erosion of the wall due to sputtering and blistering requires knowledge of the fluxes and energies of the particles which go from the plasma to the wall, of the sputtering coefficients S, and of the erosion coefficients S* for blistering. The overall erosion coefficient is equal to the sum of the sputtering coefficient and the erosion coefficient for blistering. Here the T-20 Tokamak is examined as an example of a large-scale Tokamak. 18 refs

Material migration patterns and overview of first surface analysis of the JET ITER-like wall

International Nuclear Information System (INIS)

Widdowson, A; Ayres, C F; Baron-Wiechec, A; Matthews, G F; Alves, E; Catarino, N; Brezinsek, S; Coad, J P; Likonen, J; Heinola, K; Mayer, M; Rubel, M

2014-01-01

Following the first JET ITER-like wall operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics were removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases have been drawn. The rate of deposition on inner and outer base divertor tiles and remote divertor corners was more than an order of magnitude less than during the preceding carbon wall operations, as was the concomitant deuterium retention. There was however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the mid-plane inner limiters was found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition. (paper)

Liquid Wall Chambers

Energy Technology Data Exchange (ETDEWEB)

Meier, W R

2011-02-24

The key feature of liquid wall chambers is the use of a renewable liquid layer to protect chamber structures from target emissions. Two primary options have been proposed and studied: wetted wall chambers and thick liquid wall (TLW) chambers. With wetted wall designs, a thin layer of liquid shields the structural first wall from short ranged target emissions (x-rays, ions and debris) but not neutrons. Various schemes have been proposed to establish and renew the liquid layer between shots including flow-guiding porous fabrics (e.g., Osiris, HIBALL), porous rigid structures (Prometheus) and thin film flows (KOYO). The thin liquid layer can be the tritium breeding material (e.g., flibe, PbLi, or Li) or another liquid metal such as Pb. TLWs use liquid jets injected by stationary or oscillating nozzles to form a neutronically thick layer (typically with an effective thickness of {approx}50 cm) of liquid between the target and first structural wall. In addition to absorbing short ranged emissions, the thick liquid layer degrades the neutron flux and energy reaching the first wall, typically by {approx}10 x x, so that steel walls can survive for the life of the plant ({approx}30-60 yrs). The thick liquid serves as the primary coolant and tritium breeding material (most recent designs use flibe, but the earliest concepts used Li). In essence, the TLW places the fusion blanket inside the first wall instead of behind the first wall.

Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-01-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1994-07-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Analysis of three ex-vessel loss-of-coolant accidents in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-01-01

An ex-vessel LOCA may be caused by a rupture of a cooling pipe located outside the vacuum vessel. No plasma shutdown and no other counteractions have been assumed in order to study the worst case conditions of the accidents. The next three ex-vessel LOCAs in the primary cooling system of the first wall have been analysed: 1. a large break ex-vessel LOCA caused by a rupture of the cold leg (inner diameter 0.314 m) of the main circuit; 2. an intermediate break ex-vessel LOCA caused by a rupture of a sector inlet feeder (inner diameter 0.158 m); 3. an intermediate break ex-vessel LOCA caused by a rupture of the surge line (inner diameter 0.180 m) of the pressurizer. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the first two scenarios, melting in the first wall starts about 90 s after break initiation. In the third scenario, melting in the first wall start about 323 s after break initiation. Special emphasis has been paid to the characteristics of the break flows, the transient thermal-hydraulic behaviour of the cooling system, and the temperature development in the first wall. (orig.)

Model experiments to study the first wall erosion by vacuum arcs

Energy Technology Data Exchange (ETDEWEB)

Karpov, D.A.; Saksagansky, G.L. (Leningradskij Nauchno-Issledovatel' skij Inst. (USSR). Electrophysical Apparatus); Paszti, F.; Szilagyi, E.; Manuaba, A. (Hungarian Academy of Sciences, Budapest. Central Research Inst. for Physics)

Unipolar arcs acting on the first wall of future thermonuclear reactors were modelled by bipolar arcs burning on the side surface of a cylindrical titanium cathode. Erosion rate and spatial distribution of the material sputtered in arcs were investigated by Rutherford Backscattering (RBS) analysis of collector probes. The obtianed results will be discussed as a function of arc current and the intensity of the applied vault-shaped magnetic field. (orig.).

Model experiments to study the first wall erosion by vacuum arcs

International Nuclear Information System (INIS)

Karpov, D.A.; Saksagansky, G.L.; Paszti, F.; Szilagyi, E.; Manuaba, A.

1989-01-01

Unipolar arcs acting on the first wall of future thermonuclear reactors were modelled by bipolar arcs burning on the side surface of a cylindrical titanium cathode. Erosion rate and spatial distribution of the material sputtered in arcs were investigated by Rutherford Backscattering (RBS) analysis of collector probes. The obtianed results will be discussed as a function of arc current and the intensity of the applied vault-shaped magnetic field. (orig.)

Surface segregation in binary alloy first wall candidate materials

International Nuclear Information System (INIS)

Gruen, D.M.; Krauss, A.R.; Mendelsohn, M.H.; Susman, S.; Argonne National Lab., IL

1982-01-01

We have been studying the conditions necessary to produce a self-sustaining stable lithium monolayer on a metal substrate as a means of creating a low-Z film which sputters primarily as secondary ions. It is expected that because of the toroidal field, secondary ions originating at the first wall will be returned and contribute little to the plasma impurity influx. Aluminum and copper have, because of their high thermal conductivity and low induced radioactivity, been proposed as first wall candidate materials. The mechanical properties of the pure metals are very poorly suited to structural applications and an alloy must be used to obtain adequate hardness and tensile strength. In the case of aluminum, mechanical properties suitable for aircraft manufacture are obtained by the addition of a few at% Li. In order to investigate alloys of a similar nature as candidate structural materials for fusion machines we have prepared samples of Li-doped aluminum using both a pyro-metallurgical and a vapor-diffusion technique. The sputtering properties and surface composition have been studied as a function of sample temperature and heating time, and ion beam mass. The erosion rate and secondary ion yield of both the sputtered Al and Li have been monitored by secondary ion mass spectroscopy and Auger analysis providing information on surface segregation, depth composition profiles, and diffusion rates. The surface composition ahd lithium depth profiles are compared with previously obtained computational results based on a regular solution model of segregation, while the partial sputtering yields of Al and Li are compared with results obtained with a modified version of the TRIM computer program. (orig.)

Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

International Nuclear Information System (INIS)

Kim, Geon Woo; Lee, Jeong Hun; Cho, Hyoung Kyu; Park, Goon Cherl; Im, Ki Hak

2016-01-01

The purpose of this study is to simulate thermal-hydraulic behavior of a single blanket module when plasma disruption occurs. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. The thermal design, evaluation and validation have been performed in order to establish the conceptual design guidelines of the water-cooled breeding blanket for the K-DEMO reactor. As a part of the NFRI research, Seoul National University (SNU) is conducting transient thermal-hydraulic analysis to confirm the integrity of blanket system for plasma disruption events. Vertical displacement events (VDE) with high heat flux can cause melting and vaporization of plasma facing materials (PFCs) and also burnout of coolant channels. In order to simulate melting of first wall in blanket module when VDE occurs, one-dimensional heat conduction equations were solved numerically with modification of the specific heat of the first wall materials using effective heat capacity method. Temperature profiles in first wall for VDE are shown in fig 7 - 9. At first, temperature of tungsten rapidly raised and even exceeded its melting temperature. When VDE just ended at 0.1 second, 0.83 mm thick of tungsten melted. But the other materials including vanadium and RAFM didn't exceed their melting temperatures after 500 seconds

Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon Woo; Lee, Jeong Hun; Cho, Hyoung Kyu; Park, Goon Cherl [Seoul National University, Seoul (Korea, Republic of); Im, Ki Hak [NFRI, Daejeon (Korea, Republic of)

2016-05-15

The purpose of this study is to simulate thermal-hydraulic behavior of a single blanket module when plasma disruption occurs. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. The thermal design, evaluation and validation have been performed in order to establish the conceptual design guidelines of the water-cooled breeding blanket for the K-DEMO reactor. As a part of the NFRI research, Seoul National University (SNU) is conducting transient thermal-hydraulic analysis to confirm the integrity of blanket system for plasma disruption events. Vertical displacement events (VDE) with high heat flux can cause melting and vaporization of plasma facing materials (PFCs) and also burnout of coolant channels. In order to simulate melting of first wall in blanket module when VDE occurs, one-dimensional heat conduction equations were solved numerically with modification of the specific heat of the first wall materials using effective heat capacity method. Temperature profiles in first wall for VDE are shown in fig 7 - 9. At first, temperature of tungsten rapidly raised and even exceeded its melting temperature. When VDE just ended at 0.1 second, 0.83 mm thick of tungsten melted. But the other materials including vanadium and RAFM didn't exceed their melting temperatures after 500 seconds.

Manufacturing and testing of a ITER First Wall Semi-Prototype for EUDA pre-qualification

International Nuclear Information System (INIS)

Banetta, S.; Bellin, B.; Lorenzetto, P.; Zacchia, F.; Boireau, B.; Bobin, I.; Boiffard, P.; Cottin, A.; Nogue, P.; Mitteau, R.; Eaton, R.; Raffray, R.; Bürger, A.; Du, J.; Linke, J.; Pintsuk, G.; Weber, T.

2015-01-01

Highlights: • Three ITER First Wall Small Scale Mock-ups were manufactured passing factory acceptance tests. • One of the Small Scale Mock-ups passed the thermal fatigue tests (15,000 cycles at 2 MW/m"2). • The ITER First Wall Semi-Prototype was manufactured and is being High Heat Flux tested. • Preliminary results upto 2 MW/m"2 show an overall compliance with the acceptance criteria. • Next step for EU Domestic Agency qualification is the fabrication and testing of a Full-Scale Prototype. - Abstract: This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m"2 is described in detail.

Manufacturing and testing of a ITER First Wall Semi-Prototype for EUDA pre-qualification

Energy Technology Data Exchange (ETDEWEB)

Banetta, S., E-mail: stefano.banetta@f4e.europa.eu [Fusion For Energy, Torres Diagonal Litoral, B3, Carrer Josep Pla 2, 08019 Barcelona (Spain); Bellin, B.; Lorenzetto, P.; Zacchia, F. [Fusion For Energy, Torres Diagonal Litoral, B3, Carrer Josep Pla 2, 08019 Barcelona (Spain); Boireau, B.; Bobin, I.; Boiffard, P.; Cottin, A.; Nogue, P. [AREVA NP PTCMI-F, Centre Technique, Fusion, 71200 Le Creusot (France); Mitteau, R.; Eaton, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Bürger, A.; Du, J.; Linke, J.; Pintsuk, G.; Weber, T. [Forschungszentrum Jülich, Institute of Energy and Climate Research, Jülich (Germany)

2015-10-15

Highlights: • Three ITER First Wall Small Scale Mock-ups were manufactured passing factory acceptance tests. • One of the Small Scale Mock-ups passed the thermal fatigue tests (15,000 cycles at 2 MW/m{sup 2}). • The ITER First Wall Semi-Prototype was manufactured and is being High Heat Flux tested. • Preliminary results upto 2 MW/m{sup 2} show an overall compliance with the acceptance criteria. • Next step for EU Domestic Agency qualification is the fabrication and testing of a Full-Scale Prototype. - Abstract: This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m{sup 2} is described in detail.

Development of joining processes and fabrication of US first wall qualification mockups for ITER

International Nuclear Information System (INIS)

Watson, Roger M.; Puskar, Joseph David; Ulrickson, Michael Andrew; Goods, Steven Howard

2009-01-01

We report here the fabrication processes used to manufacture US Party Team First Wall Qualification Mockups along with the detailed microstructural characterization and mechanical properties of the Be/CuCrZr/316L HIP bonds. A companion submission to this conference describes details of the PMTF heat flux testing and the performance of the first US FWQM.

Limiter and first wall of the fusion reactor blanket

International Nuclear Information System (INIS)

Danilov, I.; Skladnov, K.; Kolganov, V.

1994-01-01

Previous designing of the first wall and limiter has allowed to determine their possible embodiment depending on the parameters and operation conditions of the blanket. As a rule limiter is a separate structure located on the plasma facing surface of the blanket assembly. Possible versions of the limiter/FW which may be considered: (1) limiters with mechanical attachment of the protective part; (2) limiters with the attachment with brazing; (3) limiters with common/separate cooling system; (4) limiter as a substitute of the FW. Generally the FW/limiter structure includes protective shield and its cooling system which consist of protective coating, heat accumulator, conductive layer and attachment locks

LIFE Materials: Topical Assessment Report for LIFE Volume 1 TOPIC: Solid First Wall and Structural Components TASK: Radiation Effects on First Wall

Energy Technology Data Exchange (ETDEWEB)

Caro, A

2008-11-26

This report consists of the following chapters: CHAPTER A: LIFE Requirements for Materials. Part 1: The structure of the First Wall--Basic requirements; A qualitative view of the challenge; The candidate materials; and Base-line material's properties. CHAPTER B: Summary of Existing Knowledge--Brief historical introduction; Design window; The temperature window; Evolution of the design window with damage; Damage calculations; He and H production; Swelling resistance; Incubation dose for swelling; Design criterion No. 1, Strength; Design criterion No. 2, Corrosion resistance; Design criterion No. 3, Creep resistance; Design criterion No. 4, Radiation induced embrittlement; and Conclusions. CHAPTER C: Identification of Gaps in Knowledge & Vulnerabilities. CHAPTER D: Strategy and Future Work.

The design of the ITER first wall panels

Energy Technology Data Exchange (ETDEWEB)

Mitteau, R., E-mail: raphael.mitteau@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Calcagno, B.; Chappuis, P.; Eaton, R.; Gicquel, S. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Chen, J. [Southwestern Institute of Physics, Huangjing Road, Chengdu 610225 (China); Labusov, A. [Efremov Research Institute, 189631 St. Petersburg (Russian Federation); Martin, A.; Merola, M.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ulrickson, M. [Sandia National Laboratory, Albuquerque, NM (United States); Zacchia, F. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

2013-10-15

Highlights: • The ITER blanket is in the final stage of design completion. • Issues raised about the blanket heat loads and remote handling strategy are addressed, while integrating the in-vessel coils. • Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R and D activities. -- Abstract: The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R and D activities.

Heat deposition on the first wall due to ICRF-induced loss of fast ions in JT-60U

International Nuclear Information System (INIS)

Kusama, Y.; Tobita, K.; Kimura, H.; Hamamatsu, K.; Fujii, T.; Nemoto, M.; Saigusa, M.; Moriyama, S.; Tani, K.; Koide, Y.; Sakasai, A.; Nishitani, T.; Ushigusa, K.

1995-01-01

In JT-60U, the heat deposition on the first wall due to the ICRF-induced loss of fast ions was investigated by changing the position of the resonance layer in the ripple-trapping region. A heat spot appears on the first wall of the same major radius as the resonance layer of the ICRF waves. The broadening of the heat spot in the major radius direction is consistent with that of the resonance layer due to the Doppler broadening. The heat spot is considered to be formed by the ICRF-induced ripple-trapped loss of fast ions. Although the total ICRF-induced loss power to the heat spot is as low as 2% of the total ICRF power, the additional heat flux will become a new issue because of the localized heat deposition on the first wall. ((orig.))

Calculating the shrapnel generation and subsequent damage to first wall and optics components for the National Ignition Facility

International Nuclear Information System (INIS)

Tokheim, R.E.; Seaman, L.; Cooper, T.; Lew, B.; Curran, D.R.; Sanchez, J.; Anderson, A.; Tobin, M.

1996-01-01

This study computationally assesses the threat from shrapnel generation on the National Ignition Facility (NIF) first wall, final optics, and ultimately other target chamber components. Motion of the shrapnel is determined both by particle velocities resulting from the neutron deposition and by x-ray and ionic debris loading arising from explosion of the hohlraum. Material responses of different target area components are computed from one-dimensional and two-dimensional stress wave propagation codes. Well developed rate-dependent spall computational models are used for stainless steel spall and splitting. Severe cell distortion is accounted for in shine-shield and hohlraum-loading computations. Resulting distributions of shrapnel particles are traced to the first wall and optics and damage is estimated for candidate materials. First wall and optical material damage from shrapnel includes crater formation and associated extended cracking. 5 refs., 10 figs

Characterization of graded iron / tungsten layers for the first wall of fusion reactors

International Nuclear Information System (INIS)

Heuer, Simon

2017-01-01

The nuclear fusion has great potential to enable a CO 2 -neutral energy supply of future generations. The technical utilization of this energy source has hitherto been a challenge. In particular, high thermal loads and neutron-induced damage lead to extreme demands on the choice of materials for plasma-facing components (PFCs). These are therefore, as currently understood, made from a tungsten protective layer which is joined to a structure of low activation ferritic-martensitic (LAFM) steel. Due to the discrete transition of material properties at the LAFM-W joining zone as well as thermal loads, macroscopic stresses and plastic strains arise here. A feasible way to reduce this is to implement an intermediate layer with graded LAFM / W ratio, a so-called functional graded material (FGM). In the present work, macro-stresses and strains in the first wall of the fusion reactor DEMO are examined and evaluated by means of a finite element simulation. In this framework model components with and without graded interlayer are taken into account and the advantage of a FGM is emphasized. Parameter studies serve as a constructive guideline for the structural implementation of FGMs and components of the first wall. In addition, the feasibility of four methods (magnetron sputtering, liquid phase infiltration, modified atmospheric plasma spraying and electrodischarge sintering) with respect to the fabrication of FGMs is being studied. The resulting layers are microstructurally, thermo-physically and mechanically examined in detail. Based on this characterization and the finite element simulation, their suitability as a graded layer in the first wall of DEMO is evaluated and finally compared with alternative joining systems that are currently being tested in the research environment. [de

An overview of the development of the first wall and other principal components of a laser fusion power plant

International Nuclear Information System (INIS)

Sethian, John D.; Raffray, A. Rene; Latkowski, Jeffery; Blanchard, James P.; Snead, Lance; Renk, Timothy J.; Sharafat, Shahram

2005-01-01

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations

An overview of the development of the first wall and other principal components of a laser fusion power plant

Science.gov (United States)

Sethian, John D.; Raffray, A. Rene; Latkowski, Jeffery; Blanchard, James P.; Snead, Lance; Renk, Timothy J.; Sharafat, Shahram

2005-12-01

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

An overview of the development of the first wall and other principal components of a laser fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Sethian, John D. [Plasma Physics Division, Naval Research Laboratory, 4555 Overlook Av. SW, Washington, DC 20375 (United States)]. E-mail: sethian@this.nrl.navy.mil; Raffray, A. Rene [University of California, San Diego, La Jolla, CA 92093 (United States); Latkowski, Jeffery [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Blanchard, James P. [University of Wisconsin, Madison, WI 53706 (United States); Snead, Lance [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Renk, Timothy J. [Sandia National Laboratory, Albuquerque, NM 87185 (United States); Sharafat, Shahram [University of California, Los Angeles, Los Angeles, CA 90095 (United States)

2005-12-15

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

Towards a strategy of reliable fusion first-wall design

International Nuclear Information System (INIS)

Schultz, J.H.

1981-05-01

Fusion first walls are subject to a large number of possible failure mechanisms, including erosion due to sputtering, arcing, blistering and vaporization and crack growth due to thermal and magnetic stresses. Each of these failure mechanisms is poorly characterized and has the potential of being severe. A strategy for designing reliably in the face of great uncertainty is discussed. Topological features beneficial to reactor availability are identified. The integration of limiter pumping with rf wave launching is discussed, as a means of simplifying reactor design. The concept of a sewer limiter is introduced, as a possible long-life limiter topology. The concept of flexible armor is discussed, as a means of extending maximum life

Scope of work for evaluating the mechanical performance of EPR first wall coatings

International Nuclear Information System (INIS)

Jones, W.B.; Van Den Avyle, J.A.

1978-01-01

An outline is presented for a proposed scope of work to evaluate the mechanical performance of candidate first wall coatings for a Tokamak-type fusion reactor. The goal of the overall program is to provide an adequate coating material and recoating process which can be manufactured by currently available vendors

Refractory metal joining for first wall applications

International Nuclear Information System (INIS)

Cadden, C.H.; Odegard, B.C.

2000-01-01

The potential use of high temperature coolant (e.g. 900 deg. C He) in first wall structures would preclude the applicability of copper alloy heat sink materials and refractory metals would be potential replacements. Brazing trials were conducted in order to examine techniques to join tungsten armor to high tungsten (90-95 wt%) or molybdenum TZM heat sink materials. Palladium-, nickel- and zirconium-based filler metals were investigated using brazing temperatures ranging from 1000 deg. C to 1275 deg. C. Palladium-nickel and palladium-cobalt braze alloys were successful in producing generally sound metallurgical joints in tungsten alloy/tungsten couples, although there was an observed tendency for the pure tungsten armor material to exhibit grain boundary cracking after bonding. The zirconium- and nickel-based filler metals produced defect-containing joints, specifically cracking and porosity, respectively. The palladium-nickel braze alloy produced sound joints in the Mo TZM/tungsten couple. Substitution of a lanthanum oxide-containing, fine-grained tungsten material (for the pure tungsten) eliminated the observed tungsten grain boundary cracking

Refractory metal joining for first wall applications

Energy Technology Data Exchange (ETDEWEB)

Cadden, C.H. E-mail: chcadde@sandia.gov; Odegard, B.C

2000-12-01

The potential use of high temperature coolant (e.g. 900 deg. C He) in first wall structures would preclude the applicability of copper alloy heat sink materials and refractory metals would be potential replacements. Brazing trials were conducted in order to examine techniques to join tungsten armor to high tungsten (90-95 wt%) or molybdenum TZM heat sink materials. Palladium-, nickel- and zirconium-based filler metals were investigated using brazing temperatures ranging from 1000 deg. C to 1275 deg. C. Palladium-nickel and palladium-cobalt braze alloys were successful in producing generally sound metallurgical joints in tungsten alloy/tungsten couples, although there was an observed tendency for the pure tungsten armor material to exhibit grain boundary cracking after bonding. The zirconium- and nickel-based filler metals produced defect-containing joints, specifically cracking and porosity, respectively. The palladium-nickel braze alloy produced sound joints in the Mo TZM/tungsten couple. Substitution of a lanthanum oxide-containing, fine-grained tungsten material (for the pure tungsten) eliminated the observed tungsten grain boundary cracking.

Refractory metal joining for first wall applications

Science.gov (United States)

Cadden, C. H.; Odegard, B. C.

2000-12-01

The potential use of high temperature coolant (e.g. 900°C He) in first wall structures would preclude the applicability of copper alloy heat sink materials and refractory metals would be potential replacements. Brazing trials were conducted in order to examine techniques to join tungsten armor to high tungsten (90-95 wt%) or molybdenum TZM heat sink materials. Palladium-, nickel- and zirconium-based filler metals were investigated using brazing temperatures ranging from 1000°C to 1275°C. Palladium-nickel and palladium-cobalt braze alloys were successful in producing generally sound metallurgical joints in tungsten alloy/tungsten couples, although there was an observed tendency for the pure tungsten armor material to exhibit grain boundary cracking after bonding. The zirconium- and nickel-based filler metals produced defect-containing joints, specifically cracking and porosity, respectively. The palladium-nickel braze alloy produced sound joints in the Mo TZM/tungsten couple. Substitution of a lanthanum oxide-containing, fine-grained tungsten material (for the pure tungsten) eliminated the observed tungsten grain boundary cracking.

Beryllium for first wall, limiter and divertor - a literature survey

International Nuclear Information System (INIS)

Schuster, A.; Smid, I.; Kny, E.

1994-01-01

A survey of the topical literature on beryllium as material for plasma interactive components in future fusion devices is given. The radiation damage which can be expected as a result of the neutron irradiation from ignited tokamak plasma is discussed. The response to high heat fluxes and simulation experiments in different test facilities are referred. Another focus will be on the material properties literature data, on joining techniques and on compatibility with other materials. The performance of a beryllium coated first wall at JET is reported. Some relevant literature on other candidate materials for plasma interactive components shall be considered

Report of the study meeting on the interaction between plasma and the first wall of a fusion reactor

International Nuclear Information System (INIS)

Miyahara, Akira; Akaishi, Kenya; Kawamura, Takaichi; Kabetani, Zenzaburo; Sagara, Akio.

1978-12-01

The study meeting on the interaction between plasma and the first wall of a fusion reactor was held from July 24 to July 27, 1978. At this meeting, discussions were made on the interaction between plasma and wall and the effect of impurities. Reports on the ISS observation concerning the Mo surface as a limiter, on the measurement of sputter rate by a microbalance, on the surface roughness of the materials for the first wall at the atomic order, on the selective sputtering of binary alloys, and on the physical and chemical sputtering on the material surface of C and SiC were also presented. The research projects of the Institute of Plasma Physics and Hokkaido University were introduced. Collaboration of two groups was considered. (Kato, T.)

Simulations of fusion chamber dynamics and first wall response in a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR)

Energy Technology Data Exchange (ETDEWEB)

Qi, J.M., E-mail: qjm06@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Wang, Z., E-mail: wangz_es@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Chu, Y.Y., E-mail: chuyanyun@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Li, Z.H., E-mail: lee_march@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China)

2016-03-15

Highlights: • Z-FFR utilizes DT neutrons to drive a sub-critical fission blanket to produce energy. • A metal shell and Ar gas are employed in the fusion chamber for shock mitigation. • Massive materials can effectively mitigate the thermal heats on the chamber wall. • The W-coated Zr-alloy first wall exhibits good viability as a long-lived component. - Abstract: In a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR), the fusion target will produce enormous energy of ∼1.5 GJ per pulse at a frequency of 0.1 Hz. Almost 20% of the fusion energy yield, approximately 300 MJ, is released in forms of pulsed X-rays. To prevent the first wall from fatal damages by the intense X-rays, a thin spherical metal shell and rare Ar buffer gas are introduced to mitigate the transient X-ray bursts. Radiation hydrodynamics in the fusion chamber were investigated by MULTI-1D simulations, and the corresponding thermal and mechanical loads on the first wall were also obtained. The simulations indicated that by optimizing the design parameters of the metal shell and Ar buffer gas, peak power flux of the thermal heats on the first wall could be mitigated to less than 10{sup 4} W/cm{sup 2} within a time scale of several milliseconds, while peak overpressures of the mechanical loads varying from 0.6 to 0.7 MPa. In addition, the thermomechanical response in a W–coated Zr-alloy first wall was performed by FWDR1D calculations using the derived thermal and mechanical loads as inputs. The temperature and stress fields were analyzed, and the corresponding elastic strains were conducted for primary lifetime estimations by using the Coffin–Manson relationships of both W and Zr-alloy. It was shown that the maximum temperature rises and stresses in the first wall were less than 50 K and 130 MPa respectively, and lifetime of the first wall would be in excess of 10{sup 9} cycles. The chamber exhibits good viability as a long-lived component to sustain the Z-FFR conceptual

Wind tunnels with adapted walls for reducing wall interference

Science.gov (United States)

Ganzer, U.

1979-01-01

The basic principle of adaptable wind tunnel walls is explained. First results of an investigation carried out at the Aero-Space Institute of Berlin Technical University are presented for two dimensional flexible walls and a NACA 0012 airfoil. With five examples exhibiting very different flow conditions it is demonstrated that it is possible to reduce wall interference and to avoid blockage at transonic speeds by wall adaptation.

Application of beryllium as first wall armour for ITER primary, baffle and limiter modules

International Nuclear Information System (INIS)

Cardella, A.; Barabash, V.; Ioki, K.; Yamada, M.; Mazul, I.; Merola, M.; Strebkov, Y.

2000-01-01

During the engineering design activities of the ITER project, beryllium has been selected as the armour material for the first wall of the primary, baffle and limiter blanket modules. These components have different requirements according to their function, so the armour design and its joining technology has been developed in order to withstand different operating and loading conditions. Extensive R and D has been performed to develop, select and characterise the beryllium material and the joining techniques. In parallel, beryllium plasma spray coating has been developed, mainly as a possible in situ repair method for locally damaged areas. For the reduced technical objectives / reduced cost (RTO/RC) ITER project, it is proposed to maintain Be as the reference armour material and to optimise the manufacturing technologies in order to minimise costs. The paper presents the rationale of the design choices for the application of beryllium to the blanket first wall and gives an overview of the R and D performed and the results achieved. (orig.)

Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

International Nuclear Information System (INIS)

Swami, H.L.; Danani, C.

2015-01-01

In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

Heat transfer phenomena in the first wall of the RFX fusion experiment

International Nuclear Information System (INIS)

Oliveira Pauletti, R.M. de

1988-12-01

The thermal analysis of the first wall (FW) of the RFX machine is presented. RFX is a large fusion experiment under construction at Padua, Italy. The RFX FW is briefly described, together with the critical thermal conditions. The numerical analyses performed to predict the FW thermal behaviour are presented. 1-D and 2-D finite element models give accurate predictions of the FW temperatures and of the thermal exchanges in the machine environment. (author) [pt

NDE of explosion welded copper stainless steel first wall mock-up

International Nuclear Information System (INIS)

Taehtinen, S.; Kauppinen, P.; Jeskanen, H.; Lahdenperae, K.; Ehrnsten, U.

1997-04-01

The study showed that reflection type C-mode scanning acoustic microscope (C-SAM) and internal ultrasonic inspection (IRIS) equipment can be applied for ultrasonic examination of copper stainless steel compound structures of ITER first wall mock-ups. Explosive welding can be applied to manufacture fully bonded copper stainless steel compound plates. However, explosives can be applied only for mechanical tightening of stainless steel cooling tubes within copper plate. If metallurgical bonding between stainless steel tubes and copper plate is required Hot Isostatic Pressing (HIP) method can be applied. (orig.)

Characterization of plasma sprayed beryllium ITER first wall mockups

Energy Technology Data Exchange (ETDEWEB)

Castro, R.G.; Vaidya, R.U.; Hollis, K.J. [Los Alamos National Lab., NM (United States). Material Science and Technology Div.

1998-01-01

ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/m{sup 2} without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface. (author)

Characterization of Plasma Sprayed Beryllium ITER First Wall Mockups

International Nuclear Information System (INIS)

Castro, Richard G.; Vaidya, Rajendra U.; Hollis, Kendall J.

1997-10-01

ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/sq m without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface

Erosion simulation of first wall beryllium armour under ITER transient heat loads

Science.gov (United States)

Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

2009-04-01

The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

Erosion simulation of first wall beryllium armour under ITER transient heat loads

Energy Technology Data Exchange (ETDEWEB)

Bazylev, B. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

2009-04-30

The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

Erosion simulation of first wall beryllium armour under ITER transient heat loads

International Nuclear Information System (INIS)

Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

2009-01-01

The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

Mercury flow tests (first report). Wall friction factor measurement tests and future tests plan

International Nuclear Information System (INIS)

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Sudo, Yukio

1999-07-01

In the neutron science project at JAERI, we plan to inject a pulsed proton beam of a maximum power of 5 MW from a high intense proton accelerator into a mercury target in order to produce high energy neutrons of a magnitude of ten times or more than existing facilities. The neutrons produced by the facility will be utilized for advanced field of science such as the life sciences etc. An urgent issue in order to accomplish this project is the establishment of mercury target technology. With this in mind, a mercury experimental loop with the capacity to circulate mercury up to 15 L/min was constructed to perform thermal hydraulic tests, component tests and erosion characteristic tests. A measurement of the wall friction factor was carried out as a first step of the mercury flow tests, while testing the characteristic of components installed in the mercury loop. This report presents an outline of the mercury loop and experimental results of the wall friction factor measurement. From the wall friction factor measurement, it was made clear that the wettability of the mercury was improved with an increase of the loop operation time and at the same time the wall friction factors were increased. The measured wall friction factors were much lower than the values calculated by the Blasius equation at the beginning of the loop operation because of wall slip caused by a non-wetted condition. They agreed well with the values calculated by the Blasius equation within a deviation of 10% when the sum of the operation time increased more than 11 hours. This report also introduces technical problems with a mercury circulation and future tests plan indispensable for the development of the mercury target. (author)

Mechanical and microstructural characterization of low activation steels as first wall of nuclear fusion reactors

International Nuclear Information System (INIS)

Hernandez, M.T.; Lapena, J.; Diego, G. de; Schirra, M.

1996-01-01

Currently, the design development of fusion reactors and the possible materials to use in them are being studied in parallel. One of the most critical problems in this research is the structural materials selection for the first wall and blanket. The aim of the present work is to study three low activation alloys designed in Germany in which niobium has been substituted by tantalum or cerium. The mechanical results show that the alloys containing cerium are in the same order of the low activation materials known to date, but the tantalum doped alloy produces TaC 3 precipitation that destabilizes the matrix and provokes large microstructural changes. This causes a decrease of the mechanical properties at about 600 degree centigree. This fact makes this alloy insuitable for the first wall on fusion reactors, because the working temperature is near 550 degree centigree. (Author) 11 refs

First wall thermal stress analysis for suddenly applied heat fluxes

International Nuclear Information System (INIS)

Dalessandro, J.A.

The failure criterion for a solid first wall of an inertial confinement reactor is investigated. Analytical expressions for induced thermal stresses in a plate are given. Two materials have been chosen for this investigation: grade H-451 graphite and chemically vapor deposited (CVD) β-silicon carbide. Structural failure can be related to either the maximum compressive stress produced on the surface or the maximum tensile stress developed in the interior of the plate; however, it is shown that compressive failure would predominate. A basis for the choice of the thermal shock figure of merit, k(1 - ν) sigma/E α kappa/sup 1/2/, is identified. The result is that graphite and silicon carbide rank comparably

Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

Science.gov (United States)

Sharafat, Shahram; Ghoniem, Nasr M.; Anderson, Michael; Williams, Brian; Blanchard, Jake; Snead, Lance; HAPL Team

2005-12-01

The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential

Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

International Nuclear Information System (INIS)

Sharafat, Shahram; Ghoniem, Nasr M.; Anderson, Michael; Williams, Brian; Blanchard, Jake; Snead, Lance

2005-01-01

The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential

Melting and evaporation analysis of the first wall in a water-cooled breeding blanket module under vertical displacement event by using the MARS code

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148 Gwahak-ro, Yuseong-gu, Daejeon 34133 (Korea, Republic of)

2017-05-15

Highlights: • Material phase change of first wall was simulated for vertical displacement event. • An in-house first wall module was developed to simulate melting and evaporation. • Effective heat capacity method and evaporation model were proposed. • MARS code was proposed to predict two-phase phenomena in coolant channel. • Phase change simulation was performed by coupling MARS and in-house module. - Abstract: Plasma facing components of tokamak reactors such as ITER or the Korean fusion demonstration reactor (K-DEMO) can be subjected to damage by plasma instabilities. Plasma disruptions like vertical displacement event (VDE) with high heat flux, can cause melting and vaporization of plasma facing materials and burnout of coolant channels. In this study, to simulate melting and vaporization of the first wall in a water-cooled breeding blanket under VDE, one-dimensional heat equations were solved numerically by using an in-house first wall module, including phase change models, effective heat capacity method, and evaporation model. For thermal-hydraulics, the in-house first wall analysis module was coupled with the nuclear reactor safety analysis code, MARS, to take advantage of its prediction capability for two-phase flow and critical heat flux (CHF) occurrence. The first wall was proposed for simulation according to the conceptual design of the K-DEMO, and the heat flux of plasma disruption with a value of 600 MW/m{sup 2} for 0.1 s was applied. The phase change simulation results were analyzed in terms of the melting and evaporation thicknesses and the occurrence of CHF. The thermal integrity of the blanket first wall is discussed to confirm whether the structural material melts for the given conditions.

HELCZA-High heat flux test facility for testing ITER EU first wall components.

Czech Academy of Sciences Publication Activity Database

Prokůpek, J.; Samec, K.; Jílek, R.; Gavila, P.; Neufuss, S.; Entler, Slavomír

2017-01-01

Roč. 124, November (2017), s. 187-190 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : HELCZA * High heat flux * Electron beam testing * Test facility * Plasma facing components * First wall * Divertora Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617302818

Liquid first walls for magnetic fusion energy

International Nuclear Information System (INIS)

Moir, R.W.

1996-01-01

Liquids (∼7 neutron mean free paths thick) with certain restrictions can probably be used in magnetic fusion designs between the burning plasma and the structural materials of the plant. If this works there are a number of profound advantages: lower the cost of electricity by more than 35%; remove the need to develop first wall materials saving over 4B$ in development costs; reduce the amount and kind of wastes generated in the plant; and permit a wider choice of materials. Evaporated liquid must be efficiently ionized in an edge plasma to prevent penetrating into the burning plasma and diminishing the burn rate. The fraction of evaporated material ionized is estimated to be 0.993 for Li, 0.98 for Flibe and 0.9999 for Li 17 Pb 83 . This ionized vapor would be swept along open field lines into a remote burial chamber. The most practical systems would be those with topological open field lines on the outer surface as is the case of a field reversed configuration (FRC), a Spheromak, a Z-pinch, or a mirror machine. In a Tokamak, including the Spherical Tokamak, the field lines outside the separatrix are restricted to a small volume inside the toroidal coil making for difficulties in introducing the liquid and removing the ionized vapor

Overview of workshop on 'Evaluation of simulation techniques for radiation damage in the bulk of fusion first wall materials'

International Nuclear Information System (INIS)

Leffers, T.; Singh, B.N.; Green, W.V.; Victoria, M.

1984-05-01

The main points and the main conclusions of a workshop held June 27-30 1983 at Interlaken, Switzerland, are reported. There was general agreement among the participants that ideal simulation, providing unambiguous information about the behaviour of the first wall material, is at present out of reach. In this situation the route to follow is to use the existing simulation facilities in a concerted effort to understand the damage accumulation processes and thereby create the background for prediction or appropriate simulation of the behaviour of the first wall material. (Auth.)

Overview of Workshop on Evaluation of Simulation Techniques for Radiation Damage in the Bulk of Fusion First Wall Materials

DEFF Research Database (Denmark)

Leffers, Torben; Singh, Bachu Narain; Green, W.V.

1984-01-01

of reach. In this situation the route to follow is to use the existing simulation facilities in a concerted effort to understand the damage accumulation processes and thereby create the background for prediction or appropriate simulation of the behaviour of the first wall material.......The main points and the main conclusions of a workshop held June 27–30 1983 at Interlaken, Switzerland, are reported. There was general agreement among the participants that ideal simulation, providing unambiguous information about the behaviour of the first wall material, is at present out...

Development and evaluation of first wall materials for the National Ignition Facility

International Nuclear Information System (INIS)

Burnham, A.K.; Tobin, M.T.; Anderson, A.T.; Honea, E.C.; Skulina, K.M.; Milam, D.; Evans, M.; Rainer, F.; Gerassimenko, M.

1996-01-01

Several low-Z refractory materials are evaluated for use as the NIF first wall in terms of their cost and ability to survive laser light, target emissions and debris, as well as be cleanable and not outgas excessively. Best performers contain B, C, or both, with B 4 C being the best overall. It appears possible at this time that plasma-sprayed B 4 C can be fabricated with low enough porosity and cost to be preferred to hot-pressed B 4 C, the conservative choice

Thermo-mechanical design windows for SiC/SiC composite first wall of A-SSTR2

International Nuclear Information System (INIS)

He Kaihui; Satoshi Nishio

2002-01-01

The finite element analysis and calculation is performed for the blanket first wall made of SiC/SiC composite material for Advanced Steady-state Tokamak Reactor 2, A-SSTR2, which is now conceptually designed in Naka Fusion Research Establishment, JAERI. Comparison analysis and design window is analyzed by using the finite element code ADINA 7.4. Through 2D calculation for various geometrical configurations and sensitive material properties, a fundamental guideline for first wall and blanket design is established with respect to maximum temperature, thermal and mechanical stress for many configurations. To satisfy hydrodynamic requirement, a4d4 (the dimension of coolant channel is 4 mm x 8 mm, and the distance between neighboring channels is 4 mm) is chosen as design point for high thermal conductivity up to 50 W/m·K

Low cycle fatigue lifetime of HIP bonded Bi-metallic first wall structures of fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Hatano, Toshihisa; Sato, Satoshi; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hashimoto, Toshiyuki; Kitamura, Kazunori

1998-10-01

A HIP bonded bi-metallic panel composed of a dispersion strengthened copper (DSCu) layer and type 316L stainless steel (SS316L) cooling pipes is the reference design of the ITER first wall. To examine the fatigue lifetime of the first wall panel under cyclic mechanical loads, low cycle fatigue tests of HIP bonded bi-metallic specimens made of SS316L and DSCu were conducted with the stress ratio of -1.0 and five nominal strain range conditions ranging from 0.2 to 1.0%. Elasto-plastic analysis has also been conducted to evaluate local strain ranges under the nominal strains applied. Initial cracks were observed at the inner surface of the SS316L cooling pipes for all of the specimens tested, which was confirmed by the elasto-plastic analysis that the maximum strains of the test specimens were developed at the same locations. It was found that the HIP bonded bi-metallic test specimens had a fatigue lifetime longer than that of the SS316L raw material obtained by round bar specimens. Similarly, the fatigue lifetime of the DSCu/SS316L HIP interface was also longer than the round bar test results for the HIP joints. From these results, it has been confirmed that the bi-metallic first wall panel with built-in cooling pipes made by HIP bonding has a sufficient fatigue lifetime in comparison with the raw fatigue data of the materials, which also suggests that the fatigue lifetime evaluation has an adequate margin against fracture if it follows the design fatigue curve based on the material fatigue data. (author)

Plasma-wall interactions

International Nuclear Information System (INIS)

Behrisch, Rainer

1978-01-01

The plasma wall interactions for two extreme cases, the 'vacuum model' and the 'cold gas blanket' are outlined. As a first step for understanding the plasma wall interactions the elementary interaction processes at the first wall are identified. These are energetic ion and neutral particle trapping and release, ion and neutral backscattering, ion sputtering, desorption by ions, photons and electrons and evaporation. These processes have only recently been started to be investigated in the parameter range of interest for fusion research. The few measured data and their extrapolation into regions not yet investigated are reviewed

High emissivity TiC coatings for a first wall

International Nuclear Information System (INIS)

Groot, P.

1991-08-01

Part of the First Wall of the conceptual design of Next European Torus NET consist of radiation cooled carbon tiles. Tile temperature is determined by the optical properties of facing surfaces. Heat transfer to the 316 stainless steel structure can be improved by applying a high emissivity coating. For this purpose ceramic coatings can be applied. This paper deals with development and characterization of atmospheric and vacuum plasma sprayed titanium carbide as high emissivity coatings. Microstructural evaluation of these coatings includes X-ray diffraction and light microscopy of cross-sections. Total emissivities of vacuum and atmospheric plasma sprayed TiC coatings were measured at 525 K at PTB Braunschweig. Reflection measurements were performed at ECN Petten by using a YAG laser with wavelength 1.06 μm at room temperature. The effects of compositional differences on optical properties are discussed. (author). 9 refs.; 5 figs.; 1 tab

Shielding wall for thermonuclear device

International Nuclear Information System (INIS)

Uchida, Takaho.

1989-01-01

This invention concerns shielding walls opposing to plasmas of a thermonuclear device and it is an object thereof to conduct reactor operation with no troubles even if a portion of shielding wall tiles should be damaged. That is, the shielding wall tiles are constituted as a dual layer structure in which the lower base tiles are connected by means of bolts to first walls. Further, the upper surface tiles are bolt-connected to the layer base tiles. In this structure, the plasma thermal loads are directly received by the surface layer tiles and heat is conducted by means of conduction and radiation to the underlying base tiles and the first walls. Even upon occurrence of destruction accidents to the surface layer tiles caused by incident heat or electromagnetic force upon elimination of plasmas, since the underlying base tiles remain as they are, the first walls constituted with stainless steels, etc. are not directly exposed to the plasmas. Accordingly, the integrity of the first walls having cooling channels can be maintained and sputtering intrusion of atoms of high atom number into the plasmas can be prevented. (I.S.)

Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N.; Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N.; Lebedev, V.M.; Litunovstkii, N.V.; Mazul, I.

2007-01-01

Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm 3 . The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities ∼ 10 20 m -3 . This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material exposed to prolonged

Blanket/first wall challenges and required R&D on the pathway to DEMO

International Nuclear Information System (INIS)

Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

2015-01-01

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

Blanket/first wall challenges and required R&D on the pathway to DEMO

Energy Technology Data Exchange (ETDEWEB)

Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

2015-11-15

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

Energy Technology Data Exchange (ETDEWEB)

Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N. [A.F. IOFFE Physico-technical Institute, Russian Academy of Sciences, St Petersburg (Russian Federation); Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N. [St. Petersburg State Univ., Research Institute of Physics (Russian Federation); Lebedev, V.M. [B.P. Konstantinov Nuclear Physics Institute, Russian Academy of Science, Gatchina (Russian Federation); Litunovstkii, N.V. [D.V. Efremov Institute of Electrophysical Apparatus, St.Petersburg (Russian Federation); Mazul, I. [Development of Plasma Facing Materials and Components Laboratory, EFREMOV INSTITUTE, St Petersbourg (Russian Federation)

2007-07-01

Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm{sup 3}. The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities {approx} 10{sup 20} m{sup -3}. This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

International Nuclear Information System (INIS)

Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

1999-01-01

Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study

Turbine airfoil having near-wall cooling insert

Science.gov (United States)

Martin, Jr., Nicholas F.; Wiebe, David J.

2017-09-12

A turbine airfoil is provided with at least one insert positioned in a cavity in an airfoil interior. The insert extends along a span-wise extent of the turbine airfoil and includes first and second opposite faces. A first near-wall cooling channel is defined between the first face and a pressure sidewall of an airfoil outer wall. A second near-wall cooling channel is defined between the second face and a suction sidewall of the airfoil outer wall. The insert is configured to occupy an inactive volume in the airfoil interior so as to displace a coolant flow in the cavity toward the first and second near-wall cooling channels. A locating feature engages the insert with the outer wall for supporting the insert in position. The locating feature is configured to control flow of the coolant through the first or second near-wall cooling channel.

Conceptual Engineering Method for Attenuating He Ion Interactions on First Wall Components in the Fusion Test Facility (FTF) Employing a Low-Pressure Noble Gas

International Nuclear Information System (INIS)

Gentile, C.A.; Blanchard, W.R.; Kozub, T.; Priniski, C.; Zatz, I.; Obenschain, S.

2009-01-01

It has been shown that post detonation energetic helium ions can drastically reduce the useful life of the (dry) first wall of an IFE reactor due to the accumulation of implanted helium. For the purpose of attenuating energetic helium ions from interacting with first wall components in the Fusion Test Facility (FTF) target chamber, several concepts have been advanced. These include magnetic intervention (MI), deployment of a dynamically moving first wall, use of a sacrificial shroud, designing the target chamber large enough to mitigate the damage caused by He ions on the target chamber wall, and the use of a low pressure noble gas resident in the target chamber during pulse power operations. It is proposed that employing a low-pressure (∼ 1 torr equivalent) noble gas in the target chamber will thermalize energetic helium ions prior to interaction with the wall. The principle benefit of this concept is the simplicity of the design and the utilization of (modified) existing technologies for pumping and processing the noble ambient gas. Although the gas load in the system would be increased over other proposed methods, the use of a 'gas shield' may provide a cost effective method of greatly extending the first wall of the target chamber. An engineering study has been initiated to investigate conceptual engineering methods for implementing a viable gas shield strategy in the FTF.

Additive manufacturing of ITER first wall panel parts by two approaches: Selective laser melting and electron beam melting

International Nuclear Information System (INIS)

Zhong, Yuan; Rännar, Lars-Erik; Wikman, Stefan; Koptyug, Andrey; Liu, Leifeng; Cui, Daqing; Shen, Zhijian

2017-01-01

Highlights: • A novel way using additive manufacturing to fabricated ITER First Wall Panel parts is proposed. • ITER First Wall Panel parts successfully manufactured by both SLM and EBM are compared. • Physical and mechanical properties of SLM and EBM SS316L are clearly compared. • Problems encountered for large scale part building were discussed and possible solutions are given. - Abstract: Fabrication of ITER First Wall (FW) Panel parts by two additive manufacturing (AM) technologies, selective laser melting (SLM) and electron beam melting (EBM), was supported by Fusion for Energy (F4E). For the first time, AM is applied to manufacture ITER In-Vessel parts with complex design. Fully dense SS316L was prepared by both SLM and EBM after developing optimized laser/electron beam parameters. Characterizations on the density, magnetic permeability, microstructure, defects and inclusions were carried out. Tensile properties, Charpy-impact properties and fatigue properties of SLM and EBM SS316L were also compared. ITER FW Panel parts were successfully fabricated by both SLM and EBM in a one-step building process. The SLM part has smoother surface, better size accuracy while the EBM part takes much less time to build. Issues with removing support structures might be solved by slightly changing the design of the internal cooling system. Further investigation of the influence of neutron irradiation on materials properties between the two AM technologies is needed.

Additive manufacturing of ITER first wall panel parts by two approaches: Selective laser melting and electron beam melting

Energy Technology Data Exchange (ETDEWEB)

Zhong, Yuan [Department of Materials and Environmental Chemistry, Arrhenius Laboratory, Stockholm University, SE-106 91 Stockholm (Sweden); Rännar, Lars-Erik [Department of Quality Technology, Mechanical Engineering and Mathematics, Sports Tech Research Centre, Mid Sweden University, SE-831 25 Östersund (Sweden); Wikman, Stefan [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Koptyug, Andrey [Department of Quality Technology, Mechanical Engineering and Mathematics, Sports Tech Research Centre, Mid Sweden University, SE-831 25 Östersund (Sweden); Liu, Leifeng; Cui, Daqing [Department of Materials and Environmental Chemistry, Arrhenius Laboratory, Stockholm University, SE-106 91 Stockholm (Sweden); Shen, Zhijian, E-mail: shen@mmk.su.se [Department of Materials and Environmental Chemistry, Arrhenius Laboratory, Stockholm University, SE-106 91 Stockholm (Sweden)

2017-03-15

Highlights: • A novel way using additive manufacturing to fabricated ITER First Wall Panel parts is proposed. • ITER First Wall Panel parts successfully manufactured by both SLM and EBM are compared. • Physical and mechanical properties of SLM and EBM SS316L are clearly compared. • Problems encountered for large scale part building were discussed and possible solutions are given. - Abstract: Fabrication of ITER First Wall (FW) Panel parts by two additive manufacturing (AM) technologies, selective laser melting (SLM) and electron beam melting (EBM), was supported by Fusion for Energy (F4E). For the first time, AM is applied to manufacture ITER In-Vessel parts with complex design. Fully dense SS316L was prepared by both SLM and EBM after developing optimized laser/electron beam parameters. Characterizations on the density, magnetic permeability, microstructure, defects and inclusions were carried out. Tensile properties, Charpy-impact properties and fatigue properties of SLM and EBM SS316L were also compared. ITER FW Panel parts were successfully fabricated by both SLM and EBM in a one-step building process. The SLM part has smoother surface, better size accuracy while the EBM part takes much less time to build. Issues with removing support structures might be solved by slightly changing the design of the internal cooling system. Further investigation of the influence of neutron irradiation on materials properties between the two AM technologies is needed.

X-ray and pressure conditions on the first wall of a particle beam inertial confinement reactor

International Nuclear Information System (INIS)

Magelssen, G.R.

1979-01-01

Because of the presence of a chamber gas in a particle beam reactor cavity, nonneutron target debris created from thermonuclear burn will be modified or stopped before it reaches the first reactor wall. The resulting modified spectra and pulse lengths of the debris need to be calculated to determine first wall effects. Further, the cavity overpressure created by the momentum and energy exchange between the debris and gas must also be calculated to determine its effect. The purpose of this paper is to present results of the debris-background gas problem obtained with a one fluid, two temperature plasma hydrodynamic computer code model which includes multifrequency radiation transport. Spherical symmetry, ideal gas equation of state, and LTE for each radiation frequency group were assumed. The transport of debris ions was not included and all the debris energy was assumed to be in radiation. The calculated x-ray spectra and pulse lengths and the background overpressure are presented

Preliminary results of in situ laser-induced breakdown spectroscopy for the first wall diagnostics on EAST

Science.gov (United States)

Hu, Zhenhua; Li, Cong; Xiao, Qingmei; Liu, Ping; Fang, Ding; Mao, Hongmin; Wu, Jing; Zhao, Dongye; Ding, Hongbin; Luo, Guang-Nan; EAST Team

2017-02-01

Post-mortem methods cannot fulfill the requirement of monitoring the lifetime of the plasma facing components (PFC) and measuring the tritium inventory for the safety evaluation. Laser-induced breakdown spectroscopy (LIBS) is proposed as a promising method for the in situ study of fuel retention and impurity deposition in a tokamak. In this study, an in situ LIBS system was successfully established on EAST to investigate fuel retention and impurity deposition on the first wall without the need of removal tiles between plasma discharges. Spectral lines of D, H and impurities (Mo, Li, Si, … ) in laser-induced plasma were observed and identified within the wavelength range of 500-700 nm. Qualitative measurements such as thickness of the deposition layers, element depth profile and fuel retention on the wall are obtained by means of in situ LIBS. The results demonstrated the potential applications of LIBS for in situ characterization of fuel retention and co-deposition on the first wall of EAST. Supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB105002, 2015GB109001, and 2013GB109005), National Natural Science Foundation of China (Nos. 11575243, 11605238, 11605023), Chinesisch-Deutsches Forschungs Project (GZ765), and Korea Research Council of Fundamental Science and Technology (KRCF) under the international collaboration & research in Asian countries (PG1314).

Structure reconstruction of TiO2-based multi-wall nanotubes: first-principles calculations.

Science.gov (United States)

Bandura, A V; Evarestov, R A; Lukyanov, S I

2014-07-28

A new method of theoretical modelling of polyhedral single-walled nanotubes based on the consolidation of walls in the rolled-up multi-walled nanotubes is proposed. Molecular mechanics and ab initio quantum mechanics methods are applied to investigate the merging of walls in nanotubes constructed from the different phases of titania. The combination of two methods allows us to simulate the structures which are difficult to find only by ab initio calculations. For nanotube folding we have used (1) the 3-plane fluorite TiO2 layer; (2) the anatase (101) 6-plane layer; (3) the rutile (110) 6-plane layer; and (4) the 6-plane layer with lepidocrocite morphology. The symmetry of the resulting single-walled nanotubes is significantly lower than the symmetry of initial coaxial cylindrical double- or triple-walled nanotubes. These merged nanotubes acquire higher stability in comparison with the initial multi-walled nanotubes. The wall thickness of the merged nanotubes exceeds 1 nm and approaches the corresponding parameter of the experimental patterns. The present investigation demonstrates that the merged nanotubes can integrate the two different crystalline phases in one and the same wall structure.

Several loadings and stresses of first wall of SiC with metal liner on conceptual design of moving ring reactor 'KARIN-1'

International Nuclear Information System (INIS)

Nishikawa, Masahiro; Tachibana, Eizaburo; Watanabe, Kenji; Fujiie, Yoichi.

1983-01-01

On conceptual design of moving ring reactor ''KARIN-I'' (Output: 1850 MWe), the first wall of SiC with metal liner is considered by reason that SiC ceramics has specific features of excellent radiation damage resistance in fast neutron spectra and a very low residual radioactivity, and that the thin metal liner has good compatibility with liquid lithium and good vaccum-tight, however, a extent electromagnetic interaction. The electromagnetic force applied on the metal liner and several pressure losses of liquid lithum flow are estimated, and these forces correspond to the fluid mechanical loading on SiC first wall. Thermal loading by neutron flux is calculated on the first wall to obtain temperature distributions along the flow direction and toward the wall thickness. At the outlet of the burning section, the surface temperature of SiC rises to the value of 825 0 C on plasma side and on the metal liner, it rises to the value of 540 0 C. Finally, the stress analysis is performed. The thermal stress is about one order larger than the stress induced by the fluid mechanical loading. At the inlet of the burning section, the average tensile stress of 22.4kg/mm 2 is induced on the outer side of SiC wall, and on the inner side, the average compressive stress of -26.1kg/mm 2 is induced. At the outlet of the burning section, the tensile stress is found to oscillate between 25.5kg/mm 2 and 27.3kg/mm 2 on the outer side of SiC wall by frequency of 1 Hz, and on the inner side, the compressive stress also oscillates between -21.6kg/mm 2 and -29.0kg/mm 2 by the same frequency. These stresses are within the value of fracture stress, (72.5kg/mm 2 ). Difficult residual problems on the first wall are also discussed. (author)

Power loading on the first wall during disruptions in TFTR

International Nuclear Information System (INIS)

Janos, A.; Fredrickson, E.D.; McGuire, K.M.; Nagayama, Y.; Owens, D.K.; Wilfrid, E.

1992-01-01

Heating of the first wall of TFTR due to disruptions is investigated experimentally using an extensive array of thermocouples. By comparing results from discharges with and without disruptions, we extract effects due to the disruption alone. Disruptions preferentially heat the same areas which are heated during discharges without disruptions. Hot areas are inward protrusions or regions unshielded by neighboring areas. Peaking factors in the toroidal direction, defined as peak temperature divided by average toroidal temperature, as a function of poloidal angle, are calculated. For nondisruptive discharges, the peaking factor varies between 2 and 4. For the disruptive portion of a discharge only, the peaking factor near the midplane, where most of the energy is deposited, ranges from 3 to 5. Further away from the midplane, the peaking factor can reach 28, although the heat load is less in that region. (orig.)

Thermal shock considerations for the TFCX limiter and first wall

International Nuclear Information System (INIS)

Haines, J.R.; Fuller, G.M.

1983-01-01

Resistance to thermal shock fracture of limiter and first wall surface material candidates during plasma disruption heating conditions is evaluated. A simple, figure-of-merit type thermal shock parameter which provides a mechanism to rank material candidates is derived. Combining this figure-of-merit parameter with the parameters defining specific heating conditions yields a non-dimensional thermal shock parameter. For values of this parameter below a critical value, a given material is expected to undergo thermal shock damage. Prediction of thermal shock damage with this parameter is shown to exhibit good agreement with test data. Applying this critical parameter value approach, all materials examined in this study are expected to experience thermal shock damage for nominal TFCX plasma disruption conditions. Since the extent of this damage is not clear, tests which explore the range of expected conditions for TFCX are recommended

Characterization for fusion first-wall damage studies of using tailored D-T neutron fields

International Nuclear Information System (INIS)

Dierckx, R.; Emigh, C.R.

1979-01-01

The approximation required to apply the Bullough-Haynes results to the present calculations is somewhat crude and may imply that the details of the results contain considerable error. However, when the results for each neutron source are viewed in a relative context, several valid and important observations can be made. The almost identical swelling results obtained for the intense neutron source (INS) with a standard blanket and the fusion first wall are most striking. A further comparison with a fusion reactor shows that even the spatial and energy distributions of the neutron flux are similar. In both the INS with blanket and at the first wall of a fusion reactor, there is a radial source flux component of 14-MeV neutrons and a more or less isotropic flux component of low energy (< 14-MeV) neutrons. One must therefore conclude that from the point-of-view of neutron radiation damage, the INS with a blanket, unlike all other types of neutron sources, is not a simulation environment. It is, in fact, a small scale fusion device, and data obtained from INS irradiation experiments would represent fusion reactor results. Such data could then be used to develop correlative procedures for applying data obtained from other simulation sources to fusion reactor conditions

ITER baffle module small-scale mock-ups: first wall thermo-mechanical testing results

International Nuclear Information System (INIS)

Severi, Y.; Giancarli, L.; Poitevin, Y.; Salavy, J.F.; Le Marois, G.; Roedig, M.; Vieider, G.

1998-01-01

The EU-home team is in charge of the R and D related to the ITER baffle first wall. Five small-scale mock-ups, using Be, CFC and W tiles and different armour/heat-sink material joints under development, have been fabricated and thermomechanically tested in FE200 (Le Creusot) and JUDITH (Juelich) electron beam facilities. The small-scale mock-ups have been submitted to thermo-mechanical fatigue tests (up to failure using accelerating techniques). The objective was to determine the performances of the armour material joints under high heat flux cycles. (orig.)

Operational Windows for Dry-Wall and Wetted-Wall IFE Chambers

International Nuclear Information System (INIS)

Najmabadi, F.; Raffray, A.R.; Bromberg, L.

2004-01-01

The ARIES-IFE study was an integrated study of inertial fusion energy (IFE) chambers and chamber interfaces with the driver and target systems. Detailed analysis of various subsystems was performed parametrically to uncover key physics/technology uncertainties and to identify constraints imposed by each subsystem. In this paper, these constraints (e.g., target injection and tracking, thermal response of the first wall, and driver propagation and focusing) were combined to understand the trade-offs, to develop operational windows for chamber concepts, and to identify high-leverage research and development directions for IFE research. Some conclusions drawn in this paper are (a) the detailed characterization of the target yield and spectrum has a major impact on the chamber; (b) it is prudent to use a thin armor instead of a monolithic first wall for dry-wall concepts; (c) for dry-wall concepts with direct-drive targets, the most stringent constraint is imposed by target survival during the injection process; (d) for relatively low yield targets (<250 MJ), an operational window with no buffer gas may exist; (e) for dry-wall concepts with indirect-drive targets, a high buffer gas pressure would be necessary that may preclude propagation of the laser driver and require assisted pinch transport for the heavy-ion driver; and (f) generation and transport of aerosols in the chamber is the key feasibility issue for wetted-wall concepts

Effects of radiation and high heat flux on the performance of first-wall components. Final report

International Nuclear Information System (INIS)

Wolfer, W.G.

1985-10-01

The performance of high-heat-flux components in present and future fusion devices is strongly affected by materials properties and their changes with radiation exposure and helium content. In addition, plasma disruptions and thermal fatigue are major life-limiting aspects. A multidisciplinary approach is therefore required in the performance analysis, and the following results have been accomplished. An equation of state for helium has been derived and applied to helium bubble formation by various growth processes. Models for various radiation effects have been developed and perfected to analyze radiation-induced swelling and embrittlement for high-heat flux materials. Computer codes have been developed to predict melting, evaporation, and melt-layer stability during plasma disruptions. A structural analysis code was perfected to evaluate the stress distribution and crack propagation in a high-heat-flux component or first wall. This code was applied to a duplex structure consisting of a beryllium coating on a copper substrate. It was also used to compare the lifetimes of a first wall in a tokamak reactor made of ferritic or austenitic steel

Thermal effect of periodical bakeout on tritium inventory in first wall and permeation to coolant in reactor life

International Nuclear Information System (INIS)

Nakahara, Katsuhiko

1989-01-01

In view of safety, it is very important to control the tritium inventory in first walls and permeation to the coolant. A time-dependent diffusion and temperature calculation code, TPERM, was developed. Using this code, a numerical study on the long term effects of the bakeout temperature on tritium inventory and tritium permeation to the coolant was made. In this study, an FER type first wall (stainless steel) was considered and a cyclic operation (one cycle includes a plasma burn phase and a bakeout phase) was assumed. The results are as follows: (i) There is almost no difference in the tritium inventory in the first wall between the operation with 150 0 C-bakeout and the continuous burning operation (without bakeout). In both cases there is not tritium permeation to the coolant at 5 years' integrated burn time. The 150 0 C-bakeout is effective to release tritium in the surface (to 0.1 mm depth) region on the plasma side, but it is not effective to decrease the tritium inventory over the reactor life. (ii) To decrease the tritium inventory, a bakeout at a temperature higher than 150 0 C is necessary. But a high temperature bakeout causes earlier tritium permeation to the coolant. (iii) From these results it is suggested that the decrease the tritium inventory over the reactor life by bakeout, some form of protection against tritium permeation or a decontamination device in the cooling (or bakeout) system becomes necessary. (orig.)

Thermosyphoning analysis with the CATHENA model of the blanket and first wall cooling loop for the SEAFP reactor design

International Nuclear Information System (INIS)

Ross, W.E.

1994-02-01

This report documents the thermosyphoning analysis which was performed with the CATHENA network model of one of the blanket and first wall cooling loops of the SEAFP reactor design. This thermosyphoning analysis includes four simulations, each with a slightly different model feature or assumption. These simulations are performed to assess the primary heat transport system behaviour for a complete loss of electrical power event (total loss of flow) and to estimate the rate and extent of heat-up of the incore components. For each event, a description of some of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and outflow, and first wall and blanket temperatures is provided. (author). 4 refs., 2 tabs., 32 figs

Structural model for the first wall W-based material in ITER project

Institute of Scientific and Technical Information of China (English)

Dehua Xu; Xinkui He; Shuiquan Deng; Yong Zhao

2014-01-01

The preparation, characterization, and test of the first wall materials designed to be used in the fusion reactor have remained challenging problems in the material science. This work uses the first-principles method as implemented in the CASTEP package to study the influ-ences of the doped titanium carbide on the structural sta-bility of the W–TiC material. The calculated total energy and enthalpy have been used as criteria to judge the structural models built with consideration of symmetry. Our simulation indicates that the doped TiC tends to form its own domain up to the investigated nano-scale, which implies a possible phase separation. This result reveals the intrinsic reason for the composite nature of the W–TiC material and provides an explanation for the experimen-tally observed phase separation at the nano-scale. Our approach also sheds a light on explaining the enhancing effects of doped components on the durability, reliability, corrosion resistance, etc., in many special steels.

Recent results on high thermal energy load testing of beryllium for ITER first wall application

Science.gov (United States)

Kupriyanov, I. B.; Roedig, M.; Nikolaev, G. N.; Kurbatova, L. A.; Linke, J.; Gervash, A. A.; Giniyatulin, R. N.; Podkovyrov, V. L.; Muzichenko, A. D.; Khimchenko, L.

2011-12-01

In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m-2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m-2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m-2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm3) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m-2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

Recent results on high thermal energy load testing of beryllium for ITER first wall application

International Nuclear Information System (INIS)

Kupriyanov, I B; Nikolaev, G N; Kurbatova, L A; Roedig, M; Linke, J; Gervash, A A; Giniyatulin, R N; Podkovyrov, V L; Muzichenko, A D; Khimchenko, L

2011-01-01

In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m - 2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m - 2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m - 2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm 3 ) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m - 2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

An assessment for the erosion rate of DEMO first wall

Science.gov (United States)

Tokar, M. Z.

2018-01-01

In a fusion reactor a significant fraction of plasma particles lost from the confined volume will reach the vessel wall. The recombination of these charged species, electrons and ions of hydrogen isotopes, is a source of neutral molecules and atoms, recycling back into the plasma. Here they participate, in particular, in charge-exchange (c-x) collisions with the plasma ions and, as a result, atoms of high energies with chaotically oriented velocities are generated. A significant fraction of these hot neutrals will hit the wall, leading, as well as the outflowing fuel and impurity ions, to its erosion, limiting the reactor operation time. The rate of the wall erosion in DEMO is assessed by applying a one-dimensional model which takes into account the transport of charged and neutral species across the flux surfaces in the main part of the scrape-off layer, beyond the X-point vicinity and divertor, and by considering the shift of the centers of flux surfaces, their elongation and triangularity. Atoms generated by c-x of recycling neutrals are modeled kinetically to define firmly their energy spectrum, being of particular importance for the erosion assessment. It is demonstrated the erosion rate of the DEMO wall armor of tungsten will have a pronounced ballooning character with a significant maximum of 0.3 mm per full power year at the low field side, decreasing with an increase in the anomalous perpendicular transport in the ‘far’ SOL or the plasma density at the separatrix.

Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

Science.gov (United States)

Woolley, Robert D.

2002-01-01

A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

Inverse measurement of wall pressure field in flexible-wall wind tunnels using global wall deformation data

Science.gov (United States)

Brown, Kenneth; Brown, Julian; Patil, Mayuresh; Devenport, William

2018-02-01

The Kevlar-wall anechoic wind tunnel offers great value to the aeroacoustics research community, affording the capability to make simultaneous aeroacoustic and aerodynamic measurements. While the aeroacoustic potential of the Kevlar-wall test section is already being leveraged, the aerodynamic capability of these test sections is still to be fully realized. The flexibility of the Kevlar walls suggests the possibility that the internal test section flow may be characterized by precisely measuring small deflections of the flexible walls. Treating the Kevlar fabric walls as tensioned membranes with known pre-tension and material properties, an inverse stress problem arises where the pressure distribution over the wall is sought as a function of the measured wall deflection. Experimental wall deformations produced by the wind loading of an airfoil model are measured using digital image correlation and subsequently projected onto polynomial basis functions which have been formulated to mitigate the impact of measurement noise based on a finite-element study. Inserting analytic derivatives of the basis functions into the equilibrium relations for a membrane, full-field pressure distributions across the Kevlar walls are computed. These inversely calculated pressures, after being validated against an independent measurement technique, can then be integrated along the length of the test section to give the sectional lift of the airfoil. Notably, these first-time results are achieved with a non-contact technique and in an anechoic environment.

Effect of off-normal events on the reactor first wall

International Nuclear Information System (INIS)

Igitkhanov, Yu; Bazylev, B

2011-01-01

In this paper, we analyse the energy deposition and erosion of the W/EUROFER blanket module for the first wall (FW) of DEMO due to the runaway electrons (RE) and vertical displacements events (VDEs). The DEMO data for transients were extrapolated from ITER data by using the scaling arguments. The simulations were performed at an RE deposition energy in the range 30-100 MJ m - 2 over 0.05-0.3 s. In the case of a 'hot' VDE, all stored plasma energy is deposited on the FW area for ∼1 s. For a VDE following the thermal quench phase the remaining magnetic energy is deposited on the FW for ∼0.3 s. It is shown that the minimum W thickness needed for preventing failure of the W/EUROFER bond (assumed to be the EUROFER creep point) is large enough, causing armour melting. Both RE and VDE in DEMO will pose a major life-time issue depending on their frequency.

Effect of off-normal events on the reactor first wall

Science.gov (United States)

Igitkhanov, Yu; Bazylev, B.

2011-12-01

In this paper, we analyse the energy deposition and erosion of the W/EUROFER blanket module for the first wall (FW) of DEMO due to the runaway electrons (RE) and vertical displacements events (VDEs). The DEMO data for transients were extrapolated from ITER data by using the scaling arguments. The simulations were performed at an RE deposition energy in the range 30-100 MJ m-2 over 0.05-0.3 s. In the case of a 'hot' VDE, all stored plasma energy is deposited on the FW area for ~1 s. For a VDE following the thermal quench phase the remaining magnetic energy is deposited on the FW for ~0.3 s. It is shown that the minimum W thickness needed for preventing failure of the W/EUROFER bond (assumed to be the EUROFER creep point) is large enough, causing armour melting. Both RE and VDE in DEMO will pose a major life-time issue depending on their frequency.

Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

Energy Technology Data Exchange (ETDEWEB)

Bazylev, B., E-mail: boris.bazylev@kit.edu [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Igitkhanov, Yu.; Landman, I.; Pestchanyi, S. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

2011-10-01

Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

International Nuclear Information System (INIS)

Bazylev, B.; Igitkhanov, Yu.; Landman, I.; Pestchanyi, S.; Loarte, A.

2011-01-01

Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

Calculating the shrapnel generation and subsequent damage to first wall and optics components for the National Ignition Facility

International Nuclear Information System (INIS)

Tokheim, R.E.; Seaman, L.; Cooper, T.; Lew, B.; Curran, D.R.; Sanchez, J.; Anderson, A.; Tobin, M.

1996-01-01

The purpose of this work is to computationally assess the threat from shrapnel generation on the National Ignition Facility (NIF) first wall, final optics, and ultimately other target chamber components. Shrapnel is defined as material.that is in a solid, liquid, or clustered-vapor phase with sufficient velocity to become a threat to exposed surfaces as a consequence of its impact. Typical NIF experiments will be of two types, low neutron yield shots in which the capsule is not cryogenically cooled, and high yield shots for which cryogenic cooling of the capsule is required. For non-cryogenic shots, shrapnel would be produced by spaIIing, melting and vaporizing of ''shine shields'' by absorption and shock wave loading following 1-ω and 2-ω laser radiation. For cryogenic shots, shrapnel would be generated through shock wave splitting, spalling, and droplet formation of the cryogenic tubes following neutron energy deposition. Motion of the shrapnel is determined not only by particle velocities resulting from the neutron deposition, but also by both x-ray and debris loading arising from explosion of the hohlraum. Material responses of different target area components are computed from one- dimensional and two-dimensional stress wave propagation codes. Well developed rate-dependent spall computational models are used for stainless steel spall and splitting,. Severe cell distortion is accounted for in shine-shield and hohlraum-loading computations. Resulting distributions of shrapnel particles are traced to the first wall and optics and damage is estimated for candidate materials. First wall and optical material damage from shrapnel includes crater formation and associated extended cracking

Metastasectomy of Abdominal Wall Lesions due to Prostate Cancer Detected Through PET/CT Gallium 68-PMSA: First Case Report.

Science.gov (United States)

Ochoa, Claudia; Ramirez, Angie; Varela, Rodolfo; Godoy, Fabian; Vargas, Rafael; Forero, Jorge; Rojas, Andres; Roa, Carmen; Céspedes, Carlos; Ramos, Jose; Cabrera, Marino; Calderon, Andres

2017-05-01

Introducing the topic of abdominal wall metastasis secondary to prostate cancer with a reminder of the disease's rarity, being the first published case. This article is about a 66 year old patient diagnosed with prostate cancer [cT2aNxMx iPSA: 5,6 ng/ml Gleason 3+3, (Grade 1 Group)], treated with radical prostatectomy as well as accompanied with amplified pelvic lymphadenectomy, who subsequently presented metastatic lesions to the abdominal wall diagnosed with PET/CT Gallium 68-PMSA technique and treated with abdominal metastasectomy with adequate short term results.

Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1977-01-01

A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1978-01-01

A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

Outgassing rates before, during and after bake-out for various vacuum and first wall candidate materials of a large tokamak device

International Nuclear Information System (INIS)

Yoshikawa, H.; Gomay, J.; Sugiyama, Y.; Mizuno, M.; Komiya, S.; Tazima, T.

1977-01-01

Outgassing rates of vacuum wall candidate materials; stainless steel SS-304L and YUS-170, Inconel-625 and Hastelloy-X, and first wall materials; molybdenum, pyrolytic graphite and silicon carbide are measured before, during and after a bake-out at 500 0 C. The outgassing rate from the inside wall of the cylinder made of each material is estimated from the pressure difference between before and after a calibrated orifice. The ultimate outgassing rates of SS-304L and pyrolytic graphite, and YUS-170 Inconel-625, Hastelloy-X and molybdenum are the orders of 10 -10 and 10 -11 Pa.l.s -1 cm -2 , respectively

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

International Nuclear Information System (INIS)

Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.; Shiba, Kiyoyuki

1994-01-01

Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

Design evaluation of the semi-prototype for the ITER blanket first wall qualification

International Nuclear Information System (INIS)

Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Kim, Sun Ho; Hong, Bong Guen; Bang, In Cheol

2010-01-01

For the second qualification of the First Wall (FW) procurement of the International Thermonuclear Experimental Reactor (ITER), a semi-prototype of the FW has been designed with increased local surface heat flux up to 5 MW/m 2 . With the given conditions, the new semi-prototype design was simulated with the commercial CFD code, the ANSYS-11. The results show that the semi-prototype temperature exceeds the melting temperature of Be, and the current design is required to be modified. In order to enhance cooling, a hypervapotron was added in the design and an analysis with the same code was performed. The results show that the temperature with the hypervapotron reduced by around 100 o C but it was still higher than the melting temperature of Be. The hypervapotron mock-up was fabricated and tested with a variance of inlet coolant flow rates and heat fluxes of up to 1.75 MW/m 2 using the second Korea Heat Load Test (KoHLT-2) facility, in which heat was loaded by a graphite heater through radiation heating. Wall and coolant temperatures were measured and compared with the simulation results. So far, there is a large difference between the experiments and the simulation, and a next experiment is being prepared.

Effects of plasma disruption events on ITER first wall materials

International Nuclear Information System (INIS)

Cardella, A.; Gorenflo, H.; Lodato, A.; Ioki, K.; Raffray, R.

2000-01-01

In ITER, plasma disruption events may occur producing large fast thermal transients on plasma facing materials. Particularly important for the integrity of the first wall (FW) are relatively 'long' duration off-normal events such as plasma vertical displacement events (VDE) and runaway electrons (RE). An analytical methodology has been developed to specifically assess the effect of these events on FW plasma facing materials. For the typical energy densities and event duration expected for the primary and baffle FW, some melting and evaporation of the FW armor will occur without the beneficial effect of vapor shielding, and the metallic heat sink may also be damaged due to over-heating. The method is able to calculate the amount of melted and evaporated material, taking into account the evolution of the evaporated and melted layer and to evaluate possible effects of local temporary loss of cooling. The method has been used to analyze the effects of VDE and RE events for ITER, to study recent disruption simulation experiments and to benchmark experimental and analytical results

Production of Cu/diamond composites for first-wall heat sinks

International Nuclear Information System (INIS)

Nunes, D.; Correia, J.B.; Carvalho, P.A.; Shohoji, N.; Fernandes, H.; Silva, C.; Alves, L.C.; Hanada, K.; Osawa, E.

2011-01-01

Due to their suitable thermal conductivity and strength, copper-based materials have been considered appropriate heat sinks for first wall panels in nuclear fusion devices. However, increased thermal conductivity and mechanical strength are demanded and the concept of property tailoring involved in the design of metal matrix composites advocates for the potential of nanodiamond dispersions in copper. Copper-nanodiamond composite materials can be produced by mechanical alloying followed by a consolidation operation. Yet, this powder metallurgy route poses several challenges: nanodiamond presents intrinsically difficult bonding with copper; contamination by milling media must be closely monitored; and full densification and microstructural homogeneity should be obtained with consolidation. The present line of work is aimed at an optimization of the processing conditions of Cu-nanodiamond composites. The challenges mentioned above have been addressed, respectively, by incorporating chromium in the matrix to form a stable carbide interlayer binding the two components; by assessing the contamination originating from the milling operation through particle-induced X-ray emission spectroscopy; and by comparing the densification obtained by spark plasma sintering with hot-extrusion data from previous studies.

Surface chemistry of first wall materials - From fundamental data to modeling

International Nuclear Information System (INIS)

Linsmeier, Ch.; Reinelt, M.; Schmid, K.

2011-01-01

The application of different materials at the first wall of fusion devices, like beryllium, carbon, and tungsten in the case of ITER, unavoidably leads to the formation of compounds. These compounds are created dynamically during operation and depend on the local parameters like surface temperature, incoming particle energies and species. In dedicated, well-defined laboratory experiments, using mainly X-ray photoelectron spectroscopy and Rutherford backscattering analysis for qualitative and quantitative chemical surface analysis, the parameter space in relevant element combinations are investigated. These studies lead to a deep understanding of the reaction mechanisms under the applied conditions and to a quantitative description of reaction and diffusion processes. These data can be parameterized and integrated into a modeling approach which combines dynamic surface chemistry with the modeling of the transport in the plasma. Two different approaches for surface reaction modeling are compared and benchmarked with experimental data.

Impact of the surface quality on the thermal shock performance of beryllium armor tiles for first wall applications

Energy Technology Data Exchange (ETDEWEB)

Spilker, B., E-mail: b.spilker@fz-juelich.de; Linke, J.; Pintsuk, G.; Wirtz, M.

2016-11-01

Highlights: • Different surface qualities of S-65 beryllium are tested under high heat flux conditions. • After 1000 thermal shocks, the loaded area exhibits a crucial destruction. • Stress accelerated grain boundary oxidation/dynamic embrittlement effects are linked to the thermal shock performance of beryllium. • Thermally induced cracks form between 1 and 10 pulses and grow wider and deeper between 10 and 100 pulses. • Thermally induced cracks form and propagate independently from surface grooves and the surface quality. - Abstract: Beryllium will be applied as first wall armor material in ITER. The armor has to sustain high steady state and transient power fluxes. For transient events like edge localized modes, these transient power fluxes rise up to 1.0 GW m{sup −2} with a duration of 0.5–0.75 ms in the divertor region and a significant fraction of this power flux is deposited on the first wall as well. In the present work, the reference beryllium grade for the ITER first wall application S-65 was prepared with various surface conditions and subjected to transient power fluxes (thermal shocks) with ITER relevant loading parameters. After 1000 thermal shocks, a crucial destruction of the entire loaded area was observed and linked to the stress accelerated grain boundary oxidation (SAGBO)/dynamic embrittlement (DE) effect. Furthermore, the study revealed that the majority of the thermally induced cracks formed between 1 and 10 pulses and then grew wider and deeper with increasing pulse number. The surface quality did not influence the cracking behavior of beryllium in any detectable way. However, the polished surface demonstrated the highest resistance against the observed crucial destruction mechanism.

Metastasectomy of Abdominal Wall Lesions due to Prostate Cancer Detected Through PET/CT Gallium 68-PMSA: First Case Report

Directory of Open Access Journals (Sweden)

Claudia Ochoa

2017-05-01

Full Text Available Introducing the topic of abdominal wall metastasis secondary to prostate cancer with a reminder of the disease's rarity, being the first published case. This article is about a 66 year old patient diagnosed with prostate cancer [cT2aNxMx iPSA: 5,6 ng/ml Gleason 3+3, (Grade 1 Group], treated with radical prostatectomy as well as accompanied with amplified pelvic lymphadenectomy, who subsequently presented metastatic lesions to the abdominal wall diagnosed with PET/CT Gallium 68-PMSA technique and treated with abdominal metastasectomy with adequate short term results.

Impact of a narrow limiter SOL heat flux channel on the ITER first wall panel shaping

Czech Academy of Sciences Publication Activity Database

Kocan, M.; Pitts, R.A.; Arnoux, G.; Balboa, I.; de Vries, P.C.; Dejarnac, Renaud; Furno, I.; Goldston, R.J.; Gribov, Y.; Horáček, Jan; Komm, Michael; Labit, B.; LaBombard, B.; Lasnier, C.J.; Mitteau, R.; Nespoli, F.; Pace, D.; Pánek, Radomír; Stangeby, P.C.; Terry, J.L.; Tsui, C.; Vondráček, Petr

2015-01-01

Roč. 55, č. 3 (2015), 033019-033019 ISSN 0029-5515 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * ITER * first wall panel Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.040, year: 2015 http://iopscience.iop.org/0029-5515/55/3/033019/pdf/0029-5515_55_3_033019.pdf

Laser induced release of gases from first wall coatings for fusion applications

International Nuclear Information System (INIS)

Davis, J.W.; Haasz, A.A.; Stangeby, P.C.

1985-09-01

Wall coatings which have been produced for potential use in the JET project (Si, TiC, SiC, TiO 2 , Al 2 O 3 and MgAl 2 O 4 on Inconel 600) have been exposed to laser radiation pulses (Laser Release Analysis) in order to determine (i) the concentration of absorbed or adsorbed gases in the near surface region as a function of bakeout history, and (ii) the relative trapping behaviour of sub-eV atoms, when compared with 50-1000 eV ions. Following normal system bakeout at 500 K for 24 hours, the major species released were found to be H 2 and CO, with levels up to ∼7x10 16 H/cm 2 and ∼4x10 16 CO/cm 2 . A similar concentration of argon was found for only the TiC coating produced by sputter ion plating. A further 1-hour heating of the samples at 800-900 K resulted in a reduction of hydrogen and CO release levels by about an order of magnitude. After such preparation procedures the samples were exposed to sub-eV D 0 atoms to fluences of ∼2x10 19 D 0 /cm 2 , and deuterium retention levels were measured to be of the order of 10 14 -10 16 D/cm 2 for the various coatings. Implications of these results for JET's first-wall tritium inventory are discussed. 14 refs

The strong effect of gaps on the required shaping of the ITER first wall

International Nuclear Information System (INIS)

Stangeby, Peter

2011-01-01

Divertor tokamaks such as ITER also need limiters, namely for startup, rampdown, as well as protection of the main wall from normal and off-normal loads during the diverted phase. In future fusion devices the volume within the magnetic coils will be at a premium and it will be important to make the limiters as thin as possible. A continuous, or almost continuous, wall-limiter can be made thinner than a set of well spaced discrete limiters. The need to be able to remove and replace the components of a wall-limiter requires that its individual panels in fact be discrete but the gaps between the panels should be made as small as possible relative to the panel width to maximize the wall coverage and to minimize the extent of exposed panel edges. The modularity of a wall-limiter leads inevitably to misalignments. The gaps and misalignments reduce the power-handling capability of a modular wall-limiter relative to an ideal wall-limiter, i.e. one without any gaps or misalignments. It is shown that even small gaps and radial misalignments between the individual panels of a modular wall-limiter can require so much shaping, i.e. chamfering, of the panels in order to protect the panel edges that the peak deposited power flux density on the panel face considerably exceeds that for an ideal wall-limiter, typically by an order of magnitude. Nevertheless, compared with a set of discrete limiters which are separated by gaps larger than the limiter toroidal size, a modular, small-gap wall-limiter can still be thinner and can have lower peak deposited power flux densities (MW m -2 ), for a given total power load (MW).

Assembly & Metrology of First Wall Components of SST-1

Science.gov (United States)

Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal; Patel, Hiteshkumar; Paravastu, Yuvakiran; Jaiswal, Snehal; Chauhan, Pradeep; Babu, Gattu Ramesh; A, Arun Prakash; Bhavsar, Dhaval; Raval, Dilip C.; Khan, Ziauddin; Pradhan, Subrata

2017-04-01

First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects & procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel.

Conceptual thermal-mechanical design of the TFTR first wall armor against neutral beam impingement

International Nuclear Information System (INIS)

Chi, J.W.H.; Flaherty, R.

1976-01-01

The Tokamak Fusion Test Reactor (TFTR) is designed to operate in a pulsed mode with relatively low duty cycles. Each pulse consists of a short plasma heat-up period, a reaction period, followed by a relatively long cooldown period. Plasma heating is accomplished by ohmic heating by a current induced change in the magnetically linked ohmic heating coils, followed by neutral beam injection for further preheat and the initiation of fusion reactions. During normal operation, the bulk of the neutral beam energy will be absorbed by the plasma, while the remainder will impinge on the vacuum vessel wall. The amount of thermal energy deposited on an unprotected wall is expected to be excessive, limiting the frequency of pulses and requiring frequent wall replacement. A faulted condition would cause penetration of an unprotected wall. As a consequence, a wall armoring (or liner) concept was developed to protect the vacuum vessel wall and to permit ease of liner replacement

Local wall power loading variations in thermonuclear fusion devices

International Nuclear Information System (INIS)

Carroll, M.C.; Miley, G.H.

1989-01-01

A 2 1/2-dimensional geometric model is presented that allows calculation of power loadings at various points on the first wall of a thermonuclear fusion device. Given average wall power loadings for brems-strahlung, cyclotron radiation charged particles, and neutrons, which are determined from various plasma-physics computation models, local wall heat loads are calculated by partitioning the plasma volume and surface into cells and superimposing the heating effects of the individual cells on selected first-wall differential areas. Heat loads from the entire plasma are thus determined as a function of position on the first-wall surface. Significant differences in local power loadings were found for most fusion designs, and it was therefore concluded that the effect of local power loading variations must be taken into account when calculating temperatures and heat transfer rates in fusion device first walls

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Preliminary results

International Nuclear Information System (INIS)

Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

1993-01-01

Candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at temperatures of either 60 or 250 degrees C. Preliminary results have been obtained for several of these materials irradiated at 60 degrees C. The results show that irradiation at this temperature reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The unloading compliance technique developed for the subsize disk compact specimens works quite well, particularly for materials with lower toughness. Specimens of materials with very high toughness deform excessively, and this results in experimental difficulties

Lifetime analysis of the ITER first wall under steady-state and off-normal loads

International Nuclear Information System (INIS)

Mitteau, R; Sugihara, M; Raffray, R; Carpentier-Chouchana, S; Merola, M; Pitts, R A; Labidi, H; Stangeby, P

2011-01-01

The lifetime of the beryllium armor of the ITER first wall is evaluated for normal and off-normal operation. For the individual events considered, the lifetime spans between 930 and 35×10 6 discharges. The discrepancy between low and high estimates is caused by uncertainties about the behavior of the melt layer during off-normal events, variable plasma operation parameters and variability of the sputtering yields. These large uncertainties in beryllium armor loss estimates are a good example of the experimental nature of the ITER project and will not be truly resolved until ITER begins burning plasma operation.

Development of remote replacement system for armor tiles of first wall of FER

International Nuclear Information System (INIS)

Adachi, Junichi; Yoshizawa, Shunji; Nakano, Yasuo; Kuboyama, Takashi; Shibanuma, Kiyoshi; Kakudate, Satoshi; Oka, Kiyoshi.

1993-01-01

A remote system has been developed to replace automatically armor tiles of first walls with a single manipulator arm for the Fusion Experimental Reactor (FER). The system is composed of a manipulator arm and an end-effector (a tile replacement hand), which have a gripper of the tiles, a nutrunner to rotate attatching bolts of them and a vision sensor to measure positions of them. The system can replace the tiles by means of a visual feedback system using vision sensor, even if the positions of the tiles would have changed. As a result of tests, it has been proved that the end-effector is useful and the control system is practicable. (author)

The effect of alpha incident- and poloidal-angle distributions on blister-induced first-wall erosion

International Nuclear Information System (INIS)

Fenske, G.; Hively, L.; Miley, G.

1979-01-01

The incident velocity distribution of high-energy alpha particles bombarding the first wall of an axisymmetric tokamak is evaluated as a function of poloidal angle (theta). The resulting helium concentration profile as a function of the poloidal angle and the implant depth is calculated for a typical Experimental Power Reactor (EPR) design. The critical helium concentration for blistering is first exceeded at theta approx. 55 0 . Peak concentrations are reduced somewhat through continuous D-T sputtering which, dependent on theta, reduces the blister skin thicknesses. The blistering-induced impurity level is found to increase drastically (< approx. 50%), relative to sputtering-induced impurities, at periodic intervals corresponding to approx. 4000 hours operation when each generation of blister begins to exfoliate. (orig.)

Plasma induced material defects and threshold values for thermal loads in high temperature resistant alloys and in refractory metals for first wall application in fusion reactors

International Nuclear Information System (INIS)

Bolt, H.; Hoven, H.; Kny, E.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1986-10-01

Materials for the application in the first wall of fusion reactors of the tokamak type are subjected to pulsed heat fluxes which range from some 0.5 MW m -2 to 10 MW m -2 during normal plasma operation, and which can exceed 1000 MW m -2 during total plasma disruptions. The structural defects and material fatigue caused by this types of plasma wall interaction are investigated and the results are plotted in threshold loading curves. Additionally, the results are, as far as possible, compared with quantitative, theoretical calculations. These procedures allow a semiquantitative evaluation of the applicability of the mentioned metals in the first wall of fusion reactors. (orig.) [de

A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

International Nuclear Information System (INIS)

Birch, M.; Brocklehurst, J.E.

1987-12-01

The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)

The development of divertor and first wall armour parts at JAERI, Sandia N.L. and KFA Juelich

International Nuclear Information System (INIS)

Akiba, M.; Bolt, H.; Watson, R.; Kneringer, G.; Linke, J.

1991-01-01

The development of new armour materials, and fabrication and testings of the divertor and first wall mock-ups have worldwidely been carried out during the Conceptual Design Activites (CDA) of ITER. This paper is a review of the activities on the divertor and first wall armour components which has been performed by JAERI, Sandia National Laboratory, and KFA Juelich. The design requirements have instantly been reflected in material development. For instance, carbon fiber composites (CFCs) have already been developed to have a thermal conductivity as high as copper at room temperature. Further modification of CFC's has been made. Based on the extensive progress in armour materials, the fabrication and testings of mock-ups have been started. Divertor mock-ups which are able to endure a stationary heat flux of 8 to 15 MW/m 2 have already been developed. Some new high heat flux test facilities have been constructed and are able to simulate a heat load of plasma disruption. Extensive understanding on disruption erosion of the armour materials has been obtained by experiments with these facilities. Some mock-up tests and disruption simulating tests have been performed under international collaboration. (orig.)

Magnetic and electronic properties of single-walled Mo2C nanotube: a first-principles study

Science.gov (United States)

Jalil, Abdul; Sun, Zhongti; Wang, Dayong; Wu, Xiaojun

2018-04-01

The structural, electronic, and magnetic properties of single-walled Mo2C nanotubes are investigated by using first-principles calculations. We establish that single-walled Mo2C nanotubes can be rolled up from a graphene-like Mo2C monolayer with H- or T-type phase, i.e. H-Mo2C and T-Mo2C nanotubes. The armchair-type T-Mo2C nanotubes are more energetically stable than H-Mo2C nanotubes with the same diameter, while zigzag-type H-Mo2C nanotubes are more energetically stable than T-Mo2C nanotubes. In particular, (8, 0) H-Mo2C nanotube are more stable than Mo2C monolayer due to structural deformation. All Mo2C nanotubes are magnetic metals, independent of their chirality, and the magnetic moments of Mo atoms in the outer layer are larger than the inner. The ionic and metallic bonds in Mo2C nanotubes and delocalized electrons around Mo atoms lead to the versatile electronic and magnetic properties in them, endowing them potential applications in catalysts and electronics.

Lifetime evaluation of first wall and divertor plate by crack analyses during plasma disruptions

International Nuclear Information System (INIS)

Ohmori, Junji; Kobayashi, Takeshi; Yamada, Masao; Iida, Hiromasa

1988-05-01

The first wall and divertor armor in fusion devices are subjected to high heat and particle fluxes. In particular, disruption heating is an intense thermal shock which may cause melting or vaporization of the armor surfaces. The behavior of the armor materials is one of the major factors limiting the lifetime of these components. Generally the surface temperature of armor due to disruption gets so high that the surface may become cracked. However, even if only the surface of the armor is cracked, the function of the armor will not be lost as long as the damage is limited to within a small depth of the surface. In this study, the lifetime of the armor is evaluated by two stages: crack initiation life and crack propagation life which are related to the fatigue life and the energy release rate, respectively. Materials are graphite and C/C composite (carbon fiber reinforced carbon composite) for the first wall, and tungsten for the dinertor. For disruption conditions of Fusion Experimental Reactor, the fatigue life and the energy release rates are calculated by thermal, and stress analyses. Results show that crack initiation is expected after only a few disruptions, and the energy release rate as a function of the crack length comes up to the maximum value at a small crack length, and decreases with increasing of the crack length. This decreasing means that a crack propagation rate reduces. An unstable fracture does not occur if the maximum energy release rate does not exceed the critical energy release rate which can be obtained from the fracture toughness. (author)

Electromagnetic effects on the NET first wall caused by a plasma disruption event

International Nuclear Information System (INIS)

Crutzen, Y.R.; Biggio, M.; Farfaletti-Casali, F.

1987-01-01

During the event of a major plasma disruption, the structural components of the NET fusion reactor, such as the First Wall (FW), are subjected to strong electromagnetic transients arising from the interaction of the induced eddy currents with the large magnetic field which confines and equilibrates the plasma ring. Finite element structural analyses (static, vibration, transient dynamic) have been performed to examine stresses, deformations and reactions, generated by the electromagnetic loads, in the modular blanket-enveloping box outboard FW segment. Considering the last three engineering design variations of the outboard FW module, an improvement is obtained for the new Double Null FW configuration because of the drastic reduction of electromagnetic effects and induced stresses, mainly due to increased segmentation of the internal components

High heat load experiments for first wall materials by high power ion beams

Energy Technology Data Exchange (ETDEWEB)

Kuroda, Tsutomu; Kaneko, Osamu; Sakurai, Keiichi; Oka, Yoshihide; Shibui, Masanao; Ohmori, Junji

1985-09-01

Preliminary results are presented with some analytical calculations for thermal shock fractures of first-wall material candidates under plasma disruption heating conditions. A 120 keV - 90 A ion source has been used as an energy source to heat large specimens with heat fluxes of about 9 kW/cm/sup 2/ for pulse length of about 57 msec. Materials examined here are graphite (POCO), SiC, AlN, TiC-coated graphite, and sus 304. The SiC and AlN specimens were completely broken by only one thermal shock. The web-like surface cracks with a depth of about 0.6 mm were created in the tungsten specimen during five shots. No apparent destructive changes were observed in the graphite specimen.

Estimation of bladder wall location in ultrasound images.

Science.gov (United States)

Topper, A K; Jernigan, M E

1991-05-01

A method of automatically estimating the location of the bladder wall in ultrasound images is proposed. Obtaining this estimate is intended to be the first stage in the development of an automatic bladder volume calculation system. The first step in the bladder wall estimation scheme involves globally processing the images using standard image processing techniques to highlight the bladder wall. Separate processing sequences are required to highlight the anterior bladder wall and the posterior bladder wall. The sequence to highlight the anterior bladder wall involves Gaussian smoothing and second differencing followed by zero-crossing detection. Median filtering followed by thresholding and gradient detection is used to highlight as much of the rest of the bladder wall as was visible in the original images. Then a 'bladder wall follower'--a line follower with rules based on the characteristics of ultrasound imaging and the anatomy involved--is applied to the processed images to estimate the bladder wall location by following the portions of the bladder wall which are highlighted and filling in the missing segments. The results achieved using this scheme are presented.

Upgrade of the protection system for the first wall at JET in the ITER Be and W tiles perspective

International Nuclear Information System (INIS)

Piccolo, F.; Sartori, F.; Zabeo, L.; Conte, G.; Gauthier, E.

2006-01-01

At JET the increase of the additional heating power and the first wall upgrade with a new Be and W tiles in preparation for ITER will require improving the protection system in order to guarantee the integrity of the wall. An accurate estimation of the power load and the temperature of the tiles during a discharge will become crucial to prevent damage to the structure. In that perspective the JET protection system (WALLS) has been substantially improved and is now running at JET. The plasma magnetic information and the input power to the plasma are used to evaluate the thermal load all along the first wall. The evolution of the power distribution and tile temperature during and after a discharge are then calculated by the system. A termination of the discharge is required if a thermal limit is reached or if a vulnerable area of the vessel is exposed to an excessive level of power. An improvement in the results has been obtained using more accurate plasma boundary and magnetic information [L.Zabeo et al.'A new approach to the solution of the vacuum magnetic problem in fusion machines' this conference], developing a detailed physical model (state space) for the heat diffusion for the tiles and having a better estimation of the power deposition and distribution. The real-time data provided by the bolometry has also been taken into the account in order to evaluate the radiated power. The calibration and validation of the system have been achieved with a systematic comparison between the implemented models and the temperatures provided by the thermocouples and the new Infrared Camera. In this paper a description of the structure of the system will be briefly summarized. The models adopted to estimate the power distribution and the thermal diffusion and the comparison with IR camera will be also reported, followed by some experimental examples. (author)

Investigation of the toroidal dependence of first wall conditions in the Large Helical Device

International Nuclear Information System (INIS)

Hino, T.; Ashikawa, N.; Masuzaki, S.; Sagara, A.; Komori, A.; Yamauchi, Y.; Nobuta, Y.; Matsunaga, Y.

2010-11-01

The non-uniform wall conditions such as the fuel hydrogen retention and the erosion/deposition have been investigated in the Large Helical Device (LHD) by using toroidally and poloidally distributed material probes. They were installed in every experimental campaign from 2003 to 2010, and the evolutions of the wall conditions were clearly obtained. The wall conditions significantly depended on the operational procedures and the positions of in-vessel devices such as anodes for glow discharge and the ICRF antennas. The toroidal profiles for the amounts of retained hydrogen and helium, and the depth of wall erosion, were systematically measured. The hydrogen, helium and neon glow discharges have been conducted by using two anodes before and after the hydrogen or helium main discharges. The amount of retained hydrogen was large in the vicinity of the anodes, and drastically decreased as increase of the campaign number. This reduction well corresponds to the time period used for the hydrogen glow discharge conditioning. The erosion depth was large at the walls relatively close to the anodes, which is owing to the sputtering during the helium and neon glow discharges. The depositions of carbon and boron also depended on the positions of NBI and diborane gas inlet used for boronization, respectively. The amount of the retained helium was large at the walls close to the anodes owing to the helium glow discharge. The amount of retained helium became large at the walls close to the ICRF antennas owing to the implantation of high energy helium during the helium main discharge with the ICRF heating. In the present study, the toroidal dependences of the gas retention and the erosion/deposition in LHD were obtained, and the effects of the in-vessel devices on these plasma wall interactions were clarified. (author)

Anisotropy of domain wall resistance

Science.gov (United States)

Viret; Samson; Warin; Marty; Ott; Sondergard; Klein; Fermon

2000-10-30

The resistive effect of domain walls in FePd films with perpendicular anisotropy was studied experimentally as a function of field and temperature. The films were grown directly on MgO substrates, which induces an unusual virgin magnetic configuration composed of 60 nm wide parallel stripe domains. This allowed us to carry out the first measurements of the anisotropy of domain wall resistivity in the two configurations of current perpendicular and parallel to the walls. At 18 K, we find 8.2% and 1.3% for the domain wall magnetoresistance normalized to the wall width (8 nm) in these two respective configurations. These values are consistent with the predictions of Levy and Zhang.

Mechanical design and analysis for a EPR first wall/blanket/shield system

International Nuclear Information System (INIS)

Stevens, H.C.; Misra, B.; Youngdahl, C.K.

1978-01-01

Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are given. These developments are aimed at simplifying the design, reducing the costs and in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features

Experimental Estimation Of Energy Damping During Free Rocking Of Unreinforced Masonry Walls. First Results

International Nuclear Information System (INIS)

Sorrentino, Luigi; Masiani, Renato; Benedetti, Stefano

2008-01-01

This paper presents an ongoing experimental program on unreinforced masonry walls undergoing free rocking. Aim of the laboratory campaign is the estimation of kinetic energy damping exhibited by walls released with non-zero initial conditions of motion. Such energy damping is necessary for dynamic modelling of unreinforced masonry local mechanisms. After a brief review of the literature on this topic, the main features of the laboratory tests are presented. The program involves the experimental investigation of several parameters: 1) unit material (brick or tuff), 2) wall aspect ratio (ranging between 14.5 and 7.1), 3) restraint condition (two-sided or one-sided rocking), and 4) depth of the contact surface between facade and transverse walls (one-sided rocking only). All walls are single wythe and the mortar is pozzuolanic. The campaign is still in progress. However, it is possible to present the results on most of the mechanical properties of mortar and bricks. Moreover, a few time histories are reported, already indicating the need to correct some of the assumptions frequent in the literature

Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

International Nuclear Information System (INIS)

Herman, H.; Baker, C.C.; Maroni, V.A.

1981-10-01

The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods

Assessment of hypervapotron heat sink performance using CFD under DEMO relevant first wall conditions

Energy Technology Data Exchange (ETDEWEB)

Domalapally, Phani, E-mail: p_kumar.domalapally@cvrez.cz

2016-11-01

Highlights: • Performance of Hypervapotron heat sink was tested for First wall limiter application. • Two different materials were tested Eurofer 97 and CuCrZr at PWR conditions. • Simulations were performed to see the effect of the different inlet conditions and materials on the maximum temperature. • It was found that CuCrZr heat sink performance is far better than Eurofer heat sink at the same operating conditions. - Abstract: Among the proposed First Wall (FW) cooling concepts for European Demonstration Fusion Power Plant (DEMO), water cooled FW is one of the options. The heat flux load distribution on the FW of the DEMO reactor is not yet precisely defined. But if the heat loads on the FW are extrapolated from ITER conditions, the numbers are quite high and have to be handled none the less. The design of the FW itself is challenging as the thermal conductivity ratio of heat sink materials in ITER (CuCrZr) and in DEMO (Eurofer 97) is ∼10–12 and the operating conditions are of Pressurized Water Reactor (PWR) in DEMO instead of 70 °C and 4 MPa as in ITER. This paper analyzes the performance of Hypervapotron (HV) heat sink for FW limiter application under DEMO conditions. Where different materials, temperatures, heat fluxes and velocities are considered to predict the performance of the HV, to establish its limits in handling the heat loads before reaching the upper limits from temperature point of view. In order to assess the performance, numerical simulations are performed using commercial CFD code, which was previously validated in predicting the thermal hydraulic performance of HV geometry. Based on the results the potential usage of HV heat sink for DEMO will be assessed.

Ion bombardment effects on the fatigue life of stainless steel under simulated fusion first wall conditions

International Nuclear Information System (INIS)

Kohse, G.; Harling, O.K.

1983-01-01

Pressurized tube specimens have been exposed to simultaneous multi-energy surface ion bombardment, fast neutron irradiation and stress and temperature cycling, in a simulation of a possible fusion reactor first wall environment. After ion bombardments equivalent to months-years of reactor operation and up to 30,000 cycles, no detrimental effects on post-irradiation fatigue life were found. The ion damage is found to enhance surface cracking, but this effect is limited to the several micron surface layer in which the ions are implanted

Analysis of copper alloy to stainless steel bonded panels for ITER first wall applications

International Nuclear Information System (INIS)

Stubbins, J.F.; Kurath, P.; Drockelman, D.; Li, G.; Thomas, B.G.; Morgan, G.D.; McAfee, J.

1995-01-01

The mechanical performance of bi-layer copper alloy (Gildcop CuA115) to 316L stainless steel panels was examined. This work was to analyze potential bonding methodologies for the fabrication of ITER first wall structures, to verify the bond integrity of the fabricated panels, and to establish some mechanical performance parameters for panel structural performance. Two bonding routes were examined: explosively bonding and hot isostatically pressed (HIP) bonding. Following fabrication, the panels were mechanically loaded in tensile and fatigue tests. The mechanical performance test verified that the bond integrity was excellent, and that the primary mode of failure of the bonded panels was related to failure in the base materials rather than lack of adequate bond strength

Assembly and metrology of first wall components of SST-1

International Nuclear Information System (INIS)

Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal

2015-01-01

First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring and port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects and procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel. (author)

Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

International Nuclear Information System (INIS)

Jin Qiang; Liu Songlin; Li Min; Wang Weihua

2012-01-01

As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″ = 0.3 MW/m 2 ) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.

Thermoelectric conversion at the divertor plates and the first wall of a fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Yamaguchi, S. [National Inst. for Fusion Science, Nagoya (Japan); Sagara, A. [National Inst. for Fusion Science, Nagoya (Japan); Komori, A. [National Inst. for Fusion Science, Nagoya (Japan); Tazima, T. [National Inst. for Fusion Science, Nagoya (Japan); Motojima, O. [National Inst. for Fusion Science, Nagoya (Japan); Iiyoshi, A. [National Inst. for Fusion Science, Nagoya (Japan); Matsubara, K. [National Inst. for Fusion Science, Nagoya (Japan)]|[Yamaguchi Univ. (Japan); Onozuka, M. [National Inst. for Fusion Science, Nagoya (Japan)]|[Mitsubishi Heavy Industries Ltd. (Japan); Koganezawa, K. [National Inst. for Fusion Science, Nagoya (Japan)]|[Mitsubishi Heavy Industries Ltd. (Japan); Matsuda, T. [National Inst. for Fusion Science, Nagoya (Japan)]|[Toyo Tanso Co. Ltd. (Japan)

1995-12-31

We investigated thermoelectric conversion on the first wall and the divertor plates. Carbon, B{sub 4}C, and other carbon-based materials were tested as components of a thermoelectric element. The heat flux from the plasma was assumed to be 400 kW/m{sup 2}, and the cooling side temperature the fixed design parameter of either 350 K or 650 K. While differential radiation cooling was not considered in this study, a computer programme was used to estimate the distribution of temperature and thermal stress over the thermoelectric element. The three-legged element was conceived to be 20 cm long and 12 cm wide. The temperature in its arches reached almost 2500 K, and the maximal thermal stress was 80 MPa - still within the acceptable range for the ITER design parameter. The high thermoelectric power of B{sub 4}C accounts for the thermal efficiency of 2.8% (for 650 K) or 3.3% (for 350 K). If we find an N-type semi-conductor material with the same high absolute value as B{sub 4}C to replace carbon, the efficiency will improve to 9.4% (for 650 K) or 11% (for 350 K). Since plasma is a current-conducting medium, we discuss aspects of a plasma-connected thermoelectric element. Its efficiency would depend on the connection length of magnetic field and plasma parameters near the wall. (orig.).

First wall and divertor plate disposed facing to plasma of thermonuclear device

International Nuclear Information System (INIS)

Araki, Masanori; Suzuki, Satoshi; Akiba, Masato; Hayata, Yoshiho; Inoue, Taiji; Hayashi, Yukihiro; Kude, Yukinori

1998-01-01

In order to make the most of characteristics of each ingredient of carbon fiber-reinforced composite materials, carbon fiber unidirectionally reinforced materials and a carbon fiber three-directionally reinforced material are laminated in the direction of the thickness to form a carbon fiber-reinforced carbon composite material. In this case, the carbon fibers are continuously oriented in the direction of the thickness to constitute the carbon fiber reinforced carbon composite materials integrally. In addition, a carbon fiber-reinforced carbon composite material prepared by bonding a metal on one surface in adjacent with the unidirectional carbon fiber reinforced portion and substantially in perpendicular to the direction of the thickness of the unidirectional carbon fiber reinforced portion is used as a main constitutional material. Further, a metal tube is buried in the carbon fiber three-directionally reinforced carbon composite material. Then, a first wall and a divertor plate excellent in thermal impact resistance to be disposed facing to plasmas of a thermonuclear device can be provided. (N.H.)

Measurement of the nonaxisymmetric heat load distribution on the first wall of TFTR due to locked modes

International Nuclear Information System (INIS)

Janos, A.C.; Fredrickson, E.; McGuire, K.M.; Nagayama, Y.; Owens, D.K.

1992-01-01

The first wall of TFTR is covered in large part (23%) by an inner-wall bumper limiter which is the primary power handling structure in TFTR. The limiter is comprised of more than 2000 tiles, and is instrumented with a large number (>100) of thermocouples in a two-dimensional (2D) array, primarily for protection of the wall. While only about 5% of the tiles are monitored, this thermocouple system is nevertheless capable of mapping details in the nonaxisymmetric, as well as symmetric, heat load patterns encountered under different conditions. In particular, helical heating patterns are observed in discharges which have locked modes. The helical patterns clearly match the expected trajectories based on the m/n mode numbers obtained from Mirnov coils (m/n=2/1 and 4/1), so that the thermocouple system can and was used to identify the existence and mode number of a locked mode. While TFTR discharges rarely suffer from locked modes, locked modes always alter the heating pattern. The locked modes are found to very significantly redistribute the heat load for both ohmic and NBI heated discharges. Locked modes can make what were the coldest areas into the hottest areas, and vice versa. Locked modes also can alter the heat pattern resulting from the frequent disruptions which occur as a result of a locked mode

Ion-bombardment effects on the fatigue life of stainless steel under simulated fusion first-wall conditions

International Nuclear Information System (INIS)

Kohse, G.E.

1983-02-01

An experiment which uses the MITR-II 5 MW research reactor to simulate several aspects of the anticipated environment of a fusion reactor first wall is described. Pressurized tube specimens are subjected simultaneously to stress and temperature cycling, surface bombardment by energetic helium and lithium ions and bulk irradiation by high-energy neutrons. Analysis of the samples is aimed primarily at determining the behavior of the ion bombarded surface layer, which has a depth of 2.5 μm, with particular reference to possible effects on the fatigue life of the material

Design of a high-temperature first wall/blanket for a d-d compact Reversed-Field-Pinch reactor (CRFPR)

International Nuclear Information System (INIS)

Dabiri, A.E.; Glancy, J.E.

1983-05-01

A high-temperature first wall/blanket which would take full advantage of the absence of tritium breeding in a d-d reactor was designed. This design which produces steam at p = 7 MPa and T = 538 0 C at the blanket exit eliminates the requirement for a separate steam generator. A steam cycle with steam-to-steam reheat yielding about 37.5 percent efficiency is compatible with this design

Activities of HIP joining of plasma-facing armors in the blanket first-wall in Korea

Energy Technology Data Exchange (ETDEWEB)

Jung, Yang-Il, E-mail: yijung@kaeri.re.kr [Korea Atomic Energy Research Institute, Daedeok-daero, Daejeon 34057 (Korea, Republic of); Park, Jeong-Yong; Choi, Byoung-Kwon; Lee, Jung-Suk; Kim, Hyun-Gil; Park, Dong-Jun; Park, Jung-Hwan; Kim, Suk-Kwon; Lee, Dong-Won [Korea Atomic Energy Research Institute, Daedeok-daero, Daejeon 34057 (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Gwahak-ro, Yuseong, Daejeon 34133 (Korea, Republic of)

2016-11-01

Highlights: • HIP joints of Be/CuCrZr, Be/FMS, W/FMS were demonstrated. • The process conditions for HIP joining were developed. • For the joining of Be, coating interlayers as well as thick diffusion barrier was developed. • For the joining of W, double-staged HIP was applied for the joint integrity. • No significant defects nor a brittle failure were observed along the joint interface. - Abstract: Joining technology for dissimilar materials was developed for the fabrication of an ITER blanket first-wall, which consisted of Be, CuCrZr, and stainless steel (SS). The Be/CuCrZr/SS joint was fabricated using a hot isostatic pressing (HIP) method. Beryllium armor was joined to the CuCrZr/SS block at 580 °C under 100 MPa. The optimal interlayer coatings of Cr/Cu and Ti/Cr/Cu were developed using an ion-beam assisted physical vapor deposition. Beryllium is also a candidate armor material for the TBM first-wall. Successful joining of Be to ferritic-martensitic steel (FMS) was accomplished using an HIP method by introducing the thick diffusion barrier. A thick diffusion barrier of a Cu foil(10 μm) limited the excessive diffusion and prevented the formation of brittle phases at the Be/FMS interface. Be and FMS were bonded at 650–850 °C; however, a temperature of lower than 750 °C was recommended to avoid material degradation of FMS. In addition, the joining of W to FMS has been developed. Tungsten is another armor material applicable to more severe plasma conditions. The large difference in the thermal expansion between W and FMS was resolved by introducing the Ti interlayer and Mo separator. Moreover, the double-staged HIP (the first stage at 900 °C and 100 MPa and the second stage at 750 °C and 70 MPa) was applied to suppress the edge delamination of W/FMS joints during thermal history.

Enhancement of First Wall Damage in Iter Type Tokamak due to Lenr Effects

Science.gov (United States)

Lipson, Andrei G.; Miley, George H.; Momota, Hiromu

In recent experiments with pulsed periodic high current (J ~ 300-500 mA/cm2) D2-glow discharge at deuteron energies as low as 0.8-2.45 keV a large DD-reaction yield has been obtained. Thick target yield measurement show unusually high DD-reaction enhancement (at Ed = 1 keV the yield is about nine orders of magnitude larger than that deduced from standard Bosch and Halle extrapolation of DD-reaction cross-section to lower energies) The results obtained in these LENR experiments with glow discharge suggest nonnegligible edge plasma effects in the ITER TOKAMAK that were previously ignored. In the case of the ITER DT plasma core, we here estimate the DT reaction yield at the metal edge due to plasma ion bombardment of the first wall and/or divertor materials.

Enhancement of first wall damage in ITER type tokamak due to LENR effects

International Nuclear Information System (INIS)

Lipson, Andrei G.; Miley, George H.; Momota, Hiromu

2006-01-01

In recent experiments with pulsed periodic high current (J - 300-500 mA/cm 2 ) D 2 -glow discharge at deuteron energies as low as 0.8-2.45 keV a large DD-reaction yield has been obtained. Thick target yield measurement show unusually high DD-reaction enhancement (at E d =1 keV the yield is about nine orders of magnitude larger than that deduced from standard Bosch and Halle extrapolation of DD-reaction cross-section to lower energies). The results obtained in these LENR experiments with glow discharge suggest nonnegligible edge plasma effects in the ITER TOKAMAK that were previously ignored. In the case of the ITER DT plasma core, we here estimate the DT reaction yield at the metal edge due to plasma ion bombardment of the first wall and/or divertor materials. (author)

Fast ion power loads on ITER first wall structures in the presence of NTMs and microturbulence

International Nuclear Information System (INIS)

Kurki-Suonio, T.; Asunta, O.; Hirvijoki, E.; Koskela, T.; Snicker, A.; Sipilae, S.; Hauff, T.; Jenko, F.; Poli, E.

2011-01-01

The level and distribution of the wall power flux of energetic ions in ITER have to be known accurately in order to ensure the integrity of the first wall. Until now, most quantitative estimates have been based on the assumption that fast ion transport is dictated by neoclassical effects only. However, in ITER, the fast ion distribution is likely to be affected by various MHD effects and probably also by microturbulence. We have now upgraded our orbit-following Monte Carlo code ASCOT so that it has simple, theory-based models for neoclassical tearing mode (NTM)-type islands as well as for turbulent diffusion. ASCOT also allows for full-orbit following, which is important close to the material surfaces and, possibly, also when strong toroidal inhomogeneities are present in the magnetic field. Here we introduce the new models, preliminary results obtained with them, and how these models could be made more realistic in the future. The simulations are carried out for thermonuclear alpha particles in ITER scenario 2 plasma, because we consider this combination to be most critical for the successful operation of ITER. Neither the turbulent transport nor NTM-type islands are found to introduce alarming changes in the wall loads. However, at this stage it was not possible to combine the island structures with the non-axisymmetric magnetic field of ITER, and it remains to be seen what the combined effect of drift islands together with the toroidal ripple and local field aberrations, such as those due to test blanket modules and resonant magnetic perturbations will be.

Shear localization and effective wall friction in a wall bounded granular flow

Science.gov (United States)

Artoni, Riccardo; Richard, Patrick

2017-06-01

In this work, granular flow rheology is investigated by means of discrete numerical simulations of a torsional, cylindrical shear cell. Firstly, we focus on azimuthal velocity profiles and study the effect of (i) the confining pressure, (ii) the particle-wall friction coefficient, (iii) the rotating velocity of the bottom wall and (iv) the cell diameter. For small cell diameters, azimuthal velocity profiles are nearly auto-similar, i.e. they are almost linear with the radial coordinate. Different strain localization regimes are observed : shear can be localized at the bottom, at the top of the shear cell, or it can be even quite distributed. This behavior originates from the competition between dissipation at the sidewalls and dissipation in the bulk of the system. Then we study the effective friction at the cylindrical wall, and point out the strong link between wall friction, slip and fluctuations of forces and velocities. Even if the system is globally below the sliding threshold, force fluctuations trigger slip events, leading to a nonzero wall slip velocity and an effective wall friction coefficient different from the particle-wall one. A scaling law was found linking slip velocity, granular temperature in the main flow direction and effective friction. Our results suggest that fluctuations are an important ingredient for theories aiming to capture the interface rheology of granular materials.

Progress on the Fabrication Methods Development for the Korean Test Blanket Module First Wall in the ITER

International Nuclear Information System (INIS)

Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Cho, Seung Yon

2010-01-01

A Korean helium cooled molten lithium (HCML) test blanket module (TBM) has been designed to be tested in the International Thermonuclear Experimental Reactor (ITER) TBM and related fabrication methods have been developed especially for the purpose of joining. Since the first wall (FW) of the HCML TBM is composed of a beryllium (Be) as an armor material and a FMS as a structural one, joining with Be to FMS and FMS to FMS should be developed in order to fabricate it

Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

1981-01-01

Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

1981-01-01

Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

1981-01-01

Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Erosion simulation of first wall beryllium armour under ITER transient heat loads

Energy Technology Data Exchange (ETDEWEB)

Bazylev, B.; Janeschitz, G. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Karlsruhe (Germany); Loarte, A. [EFDA Close Support Unit Garching, Garching bei Munchen(Germany)

2007-07-01

Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10{sup 4} ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m{sup 2} in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m{sup 2} and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m{sup 2} in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be

Erosion simulation of first wall beryllium armour under ITER transient heat loads

International Nuclear Information System (INIS)

Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

2007-01-01

Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10 4 ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m 2 in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m 2 and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m 2 in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be FW macro

Assessment of thermo-mechanical behavior in CLAM steel first wall structures

International Nuclear Information System (INIS)

Liu Fubin; Yao Man

2012-01-01

Highlights: ► China Low Activation Martensitic steel (CLAM) as FW the structural material. ► The thermo-mechanical behavior of the FW was analyzed under the condition of normal ITER operation combined effect of plasma heat flux and neutron heating. ► The temperature dependence of the material physical properties of CLAM is summarized. - Abstract: The temperature and strain distributions of the mockup with distinct structural material (SS316L or China Low Activation Martensitic steel (CLAM)) in two-dimensional model were calculated and analyzed, based on a high heat flux (HHF) test recently reported with heat flux of 3.2 MW/m 2 . The calculated temperature and strain results in the first wall (FW), in which SS316L is as the structural material, showed good agreement with HHF test. By substituting CLAM steel for SS316L the contrast analysis indicates that the thermo-mechanical property for CLAM steel is better than that of SS316 at the same condition. Furthermore, the thermo-mechanical behavior of the FW was analyzed under the condition of normal ITER operation combined effect of plasma heat flux and neutron heating.

Assessment of thermo-mechanical behavior in CLAM steel first wall structures

Energy Technology Data Exchange (ETDEWEB)

Liu Fubin, E-mail: liufubin_1216@126.com [School of Materials Science and Engineering, Dalian University of Technology, Dalian 116024, Liaoning (China); Yao Man, E-mail: yaoman@dlut.edu.cn [School of Materials Science and Engineering, Dalian University of Technology, Dalian 116024, Liaoning (China)

2012-01-15

Highlights: Black-Right-Pointing-Pointer China Low Activation Martensitic steel (CLAM) as FW the structural material. Black-Right-Pointing-Pointer The thermo-mechanical behavior of the FW was analyzed under the condition of normal ITER operation combined effect of plasma heat flux and neutron heating. Black-Right-Pointing-Pointer The temperature dependence of the material physical properties of CLAM is summarized. - Abstract: The temperature and strain distributions of the mockup with distinct structural material (SS316L or China Low Activation Martensitic steel (CLAM)) in two-dimensional model were calculated and analyzed, based on a high heat flux (HHF) test recently reported with heat flux of 3.2 MW/m{sup 2}. The calculated temperature and strain results in the first wall (FW), in which SS316L is as the structural material, showed good agreement with HHF test. By substituting CLAM steel for SS316L the contrast analysis indicates that the thermo-mechanical property for CLAM steel is better than that of SS316 at the same condition. Furthermore, the thermo-mechanical behavior of the FW was analyzed under the condition of normal ITER operation combined effect of plasma heat flux and neutron heating.

Granular packings with moving side walls

International Nuclear Information System (INIS)

Landry, James W.; Grest, Gary Stephen

2004-01-01

The effects of movement of the side walls of a confined granular packing are studied by discrete element, molecular dynamics simulations. The dynamical evolution of the stress is studied as a function of wall movement both in the direction of gravity as well as opposite to it. For all wall velocities explored, the stress in the final state of the system after wall movement is fundamentally different from the original state obtained by pouring particles into the container and letting them settle under the influence of gravity. The original packing possesses a hydrostaticlike region at the top of the container which crosses over to a depth-independent stress. As the walls are moved in the direction opposite to gravity, the saturation stress first reaches a minimum value independent of the wall velocity, then increases to a steady-state value dependent on the wall velocity. After wall movement ceases and the packing reaches equilibrium, the stress profile fits the classic Janssen form for high wall velocities, while some deviations remain for low wall velocities. The wall movement greatly increases the number of particle-wall and particle-particle forces at the Coulomb criterion. Varying the wall velocity has only small effects on the particle structure of the final packing so long as the walls travel a similar distance.

Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

Energy Technology Data Exchange (ETDEWEB)

Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Garching (Germany); Chen, Yuming; Ilić, Milica; Schwab, Florian [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sieglin, Bernhard [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [EUROfusion – Programme Management Unit, Garching (Germany)

2016-11-01

Highlights: • Existing first wall designs and expected plasma heat loads are reviewed. • Heat transfer enhancement methods are investigated by CFD. • The results for heat transfer and friction are given, compared and explained. • Relations for needed pumping power and gained thermal heat are shown. • A range for the maximum permissible heat loads from the plasma is estimated. - Abstract: The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

Low-rise shear wall failure modes

International Nuclear Information System (INIS)

Farrar, C.R.; Hashimoto, P.S.; Reed, J.W.

1991-01-01

A summary of the data that are available concerning the structural response of low-rise shear walls is presented. This data will be used to address two failure modes associated with the shear wall structures. First, data concerning the seismic capacity of the shear walls with emphasis on excessive deformations that can cause equipment failure are examined. Second, data concerning the dynamic properties of shear walls (stiffness and damping) that are necessary to compute the seismic inputs to attached equipment are summarized. This case addresses the failure of equipment when the structure remains functional. 23 refs

Hydrodynamics of ultra-relativistic bubble walls

Energy Technology Data Exchange (ETDEWEB)

Leitao, Leonardo, E-mail: lleitao@mdp.edu.ar; Mégevand, Ariel, E-mail: megevand@mdp.edu.ar

2016-04-15

In cosmological first-order phase transitions, gravitational waves are generated by the collisions of bubble walls and by the bulk motions caused in the fluid. A sizeable signal may result from fast-moving walls. In this work we study the hydrodynamics associated to the fastest propagation modes, namely, ultra-relativistic detonations and runaway solutions. We compute the energy injected by the phase transition into the fluid and the energy which accumulates in the bubble walls. We provide analytic approximations and fits as functions of the net force acting on the wall, which can be readily evaluated for specific models. We also study the back-reaction of hydrodynamics on the wall motion, and we discuss the extrapolation of the friction force away from the ultra-relativistic limit. We use these results to estimate the gravitational wave signal from detonations and runaway walls.

Development of a copper alloy to beryllium HIP bonding technology for the ITER first wall

International Nuclear Information System (INIS)

Sherlock, P.; Peacock, A.T.; Mc Callum, A.D.

2005-01-01

The primary first wall (PFW) panels of the ITER blanket concept comprise a bi-metallic copper alloy/stainless steel water-cooled heatsink faced with a plasma facing material. Precipitation strengthened CuCrZr is one option for the copper alloy of the heatsink; beryllium, in the form of tiles is an option for the plasma facing material. Over recent years, the technology needed to HIP bond the beryllium tiles to CuCrZr alloy has been developed. This paper describes small samples and larger mock-ups produced during the development of this HIP bonding technology and outlines how structural analyses were used to gain an understanding of the bonding process and refine the design

In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions

Science.gov (United States)

Litunovsky, N.; Gervash, A.; Lorenzetto, P.; Mazul, I.; Melder, R.

2009-04-01

The paper describes the experimental technique and preliminary results of thermal fatigue testing of ITER first wall (FW) water-cooled mock-ups inside the core of the RBT-6 experimental fission reactor (RIAR, Dimitrovgrad, Russia). This experiment has provided simultaneous effect of neutron fluence and thermal cycling damages on the mock-ups. A PC-controlled high-temperature graphite ohmic heater was applied to provide cyclic thermal load onto the mock-ups surface. This experiment lasted for 309 effective irradiation days with a final damage level (CuCrZr) of 1 dpa in the mock-ups. About 3700 thermal cycles with a heat flux of 0.4-0.5 MW/m 2 onto the mock-ups were realized before the heater fails. Then, irradiation was continued in a non-cycling mode.

First-principles calculations on double-walled inorganic nanotubes with hexagonal chiralities

International Nuclear Information System (INIS)

Zhukovskii, Yuri F; Evarestov, Robert A; Bandura, Andrei V; Losev, Maxim V

2011-01-01

The two sets of commensurate double-walled boron nitride and titania hexagonally-structured nanotubes (DW BN and TiO 2 NTs) possessing either armchair- or zigzag-type chiralities have been considered, i.e., (n 1 ,n 1 )-(n 2 ,n 2 ) or (n 1 ,0)-(n 2 ,0), respectively. For symmetry analysis of these nanotubes, the line symmetry groups for one-periodic (1D) nanostructures with rotohelical symmetry have been applied. To analyze the structural and electronic properties of hexagonal DW NTs, a series of large-scale ab initio DFT-LCAO calculations have been performed using the hybrid Hartree-Fock/Kohn-Sham exchange-correlation functional PBE0 (as implemented in CRYSTAL-09 code). To establish the optimal inter-shell distances within DW NTs corresponding to the minima of calculated total energy, the chiral indices n 1 and n 2 of the constituent single-walled (SW) nanotubes have been successively varied.

Vibrotactile Vest and The Humming Wall

DEFF Research Database (Denmark)

Morrison, Ann; Manresa-Yee, Cristina; Knoche, Hendrik

2015-01-01

Vibrotactile information can be used to elicit sensations and encourage particular user body movements. We designed a vibrotactile vest with physiological monitoring that interacts with a vibroacoustic urban environment, The Humming Wall. In this paper, we describe the first field trial with the ......Vibrotactile information can be used to elicit sensations and encourage particular user body movements. We designed a vibrotactile vest with physiological monitoring that interacts with a vibroacoustic urban environment, The Humming Wall. In this paper, we describe the first field trial...... with the system held over a 5-week period in an urban park. We depict the participants’ experience, engagement and impressions while wearing the vibrotactile vest and interacting with the wall. We contribute with positive responses to novel interactions between the responsive environment and the vibrotactile vest...

A unified wall function for compressible turbulence modelling

Science.gov (United States)

Ong, K. C.; Chan, A.

2018-05-01

Turbulence modelling near the wall often requires a high mesh density clustered around the wall and the first cells adjacent to the wall to be placed in the viscous sublayer. As a result, the numerical stability is constrained by the smallest cell size and hence requires high computational overhead. In the present study, a unified wall function is developed which is valid for viscous sublayer, buffer sublayer and inertial sublayer, as well as including effects of compressibility, heat transfer and pressure gradient. The resulting wall function applies to compressible turbulence modelling for both isothermal and adiabatic wall boundary conditions with the non-zero pressure gradient. Two simple wall function algorithms are implemented for practical computation of isothermal and adiabatic wall boundary conditions. The numerical results show that the wall function evaluates the wall shear stress and turbulent quantities of wall adjacent cells at wide range of non-dimensional wall distance and alleviate the number and size of cells required.

Hydrodynamics of ultra-relativistic bubble walls

Directory of Open Access Journals (Sweden)

Leonardo Leitao

2016-04-01

Full Text Available In cosmological first-order phase transitions, gravitational waves are generated by the collisions of bubble walls and by the bulk motions caused in the fluid. A sizeable signal may result from fast-moving walls. In this work we study the hydrodynamics associated to the fastest propagation modes, namely, ultra-relativistic detonations and runaway solutions. We compute the energy injected by the phase transition into the fluid and the energy which accumulates in the bubble walls. We provide analytic approximations and fits as functions of the net force acting on the wall, which can be readily evaluated for specific models. We also study the back-reaction of hydrodynamics on the wall motion, and we discuss the extrapolation of the friction force away from the ultra-relativistic limit. We use these results to estimate the gravitational wave signal from detonations and runaway walls.

Failure modes of low-rise shear walls

International Nuclear Information System (INIS)

Farrar, C.R.; Reed, J.W.; Salmon, M.W.

1993-01-01

A summary of available data concerning the structural response of low-rise shear walls is presented. These data will be used to address two failure modes associated with shear wall structures. First, the data concerning the seismic capacity of the shear walls are examined, with emphasis on excessive deformations that can cause equipment failure. Second, the data concerning the dynamic properties of shear walls (stiffness and damping) that are necessary for computing the seismic inputs to attached equipment are summarized. This case addresses the failure of equipment when the structure remains functional

Cell Wall Remodeling Enzymes Modulate Fungal Cell Wall Elasticity and Osmotic Stress Resistance.

Science.gov (United States)

Ene, Iuliana V; Walker, Louise A; Schiavone, Marion; Lee, Keunsook K; Martin-Yken, Hélène; Dague, Etienne; Gow, Neil A R; Munro, Carol A; Brown, Alistair J P

2015-07-28

The fungal cell wall confers cell morphology and protection against environmental insults. For fungal pathogens, the cell wall is a key immunological modulator and an ideal therapeutic target. Yeast cell walls possess an inner matrix of interlinked β-glucan and chitin that is thought to provide tensile strength and rigidity. Yeast cells remodel their walls over time in response to environmental change, a process controlled by evolutionarily conserved stress (Hog1) and cell integrity (Mkc1, Cek1) signaling pathways. These mitogen-activated protein kinase (MAPK) pathways modulate cell wall gene expression, leading to the construction of a new, modified cell wall. We show that the cell wall is not rigid but elastic, displaying rapid structural realignments that impact survival following osmotic shock. Lactate-grown Candida albicans cells are more resistant to hyperosmotic shock than glucose-grown cells. We show that this elevated resistance is not dependent on Hog1 or Mkc1 signaling and that most cell death occurs within 10 min of osmotic shock. Sudden decreases in cell volume drive rapid increases in cell wall thickness. The elevated stress resistance of lactate-grown cells correlates with reduced cell wall elasticity, reflected in slower changes in cell volume following hyperosmotic shock. The cell wall elasticity of lactate-grown cells is increased by a triple mutation that inactivates the Crh family of cell wall cross-linking enzymes, leading to increased sensitivity to hyperosmotic shock. Overexpressing Crh family members in glucose-grown cells reduces cell wall elasticity, providing partial protection against hyperosmotic shock. These changes correlate with structural realignment of the cell wall and with the ability of cells to withstand osmotic shock. The C. albicans cell wall is the first line of defense against external insults, the site of immune recognition by the host, and an attractive target for antifungal therapy. Its tensile strength is conferred by

Analysis of loss of electrical power with the CATHENA model of the blanket and first wall cooling loop for the SEAFP reactor design

International Nuclear Information System (INIS)

Ross, W.E.

1994-08-01

This report documents the thermosyphoning analysis which was performed with the CATHENA network model of one of the blanket and first wall cooling loops of the SEAFP reactor design. This thermosyphoning analysis is similar to that reported in CFFTP-G--9355, Volume 4 except that a much larger decay power transient is used. Also, the pressurizer heaters are turned off following the loss of electrical power. This analysis is performed to assess the primary heat transport system behaviour for a complete loss of electrical power event (total loss of flow) and to estimate the rate of heatup of the in-core components. A description of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and steam bleed flow, and first wall and blanket temperatures are provided. (author). 8 refs., 2 tabs., 26 figs

Preliminary electromagnetic, thermal and mechanical design for first wall and vacuum vessel of FAST

Energy Technology Data Exchange (ETDEWEB)

Lucca, F., E-mail: Flavio.Lucca@LTCalcoli.it [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Bertolini, C. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Crescenzi, F.; Crisanti, F. [C.R. ENEA Frascati – UT FUS, Via E. Fermi 45, IT-00044 Frascati, RM (Italy); Di Gironimo, G. [CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Labate, C. [CREATE, Università di Napoli Parthenope, Via Acton 38, 80133 Napoli (Italy); Manzoni, M.; Marconi, M.; Pagani, I. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Ramogida, G. [C.R. ENEA Frascati – UT FUS, Via E. Fermi 45, IT-00044 Frascati, RM (Italy); Renno, F. [CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Roccella, M. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Roccella, S. [C.R. ENEA Frascati – UT FUS, Via E. Fermi 45, IT-00044 Frascati, RM (Italy); Viganò, F. [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy)

2015-10-15

The fusion advanced study torus (FAST), with its compact design, high toroidal field and plasma current, faces many of the problems met by ITER, and at the same time anticipates much of the DEMO relevant physics and technology. The conceptual design of the first wall (FW) and the vacuum vessel (VV) has been defined on the basis of FAST operative conditions and of “Snow Flakes” (SF) magnetic topology, which is also relevant for DEMO. The EM loads are one of the most critical load components for the FW and the VV during plasma disruptions and a first dimensioning of these components for such loads is mandatory. During this first phase of R&D activities the conceptual design of the FW and VV have been assessed estimating, by means of FE simulations, the EM loads due to a typical vertical disruption event (VDE) in FAST. EM loads were then transferred on a FE mechanical model of the FAST structures and the mechanical response of the FW and VV design for the analyzed VDE event was assessed. The results indicate that design criteria are not fully satisfied by the current drawing of the VV and FW components. The most critical regions have been individuated and the effect of some geometrical and material changes has been checked in order to improve the structure.

Status of the beryllium tile bonding qualification activities for the manufacturing of the ITER first wall

International Nuclear Information System (INIS)

Mitteau, Raphaël; Eaton, R.; Perez, G.; Zacchia, F.; Banetta, S.; Bellin, B.; Gervash, A.; Glazunov, D.; Chen, J.

2015-01-01

The preparation of the manufacturing of the ITER first wall involves a qualification stage. The qualification aims at demonstrating that manufacturers can deliver the needed reliability and quality for the beryllium to copper bond, before the manufacturing can commence. The qualification is done on semi-prototype, containing relevant features relative to the beryllium armour (about 1/6 of the panel size). The qualification is done by the participating parties, firstly by a manufacturing semi-prototype and then by testing it under heat flux. One semi-prototype is manufactured and is being tested, and further from other manufacturers are still to come. The qualification programme is accompanied by bond defect investigations, which aim at defining defect acceptance criteria. Qualification and defect acceptance programme are supported by thermal and stress analyses, with good agreement regarding the thermal results, and some insights about the governing factors to bond damage.

Status of the beryllium tile bonding qualification activities for the manufacturing of the ITER first wall

Energy Technology Data Exchange (ETDEWEB)

Mitteau, Raphaël, E-mail: Raphael.mitteau@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Eaton, R.; Perez, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Zacchia, F.; Banetta, S.; Bellin, B. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Gervash, A.; Glazunov, D. [Efremov Research Institute, 189631 St. Petersburg (Russian Federation); Chen, J. [Southwestern Institute of Physics, Huangjing Road, Chengdu 610225 (China)

2015-10-15

The preparation of the manufacturing of the ITER first wall involves a qualification stage. The qualification aims at demonstrating that manufacturers can deliver the needed reliability and quality for the beryllium to copper bond, before the manufacturing can commence. The qualification is done on semi-prototype, containing relevant features relative to the beryllium armour (about 1/6 of the panel size). The qualification is done by the participating parties, firstly by a manufacturing semi-prototype and then by testing it under heat flux. One semi-prototype is manufactured and is being tested, and further from other manufacturers are still to come. The qualification programme is accompanied by bond defect investigations, which aim at defining defect acceptance criteria. Qualification and defect acceptance programme are supported by thermal and stress analyses, with good agreement regarding the thermal results, and some insights about the governing factors to bond damage.

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

International Nuclear Information System (INIS)

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

Conditions of vacuum physics for selection of the material of first wall and diaphragm of the demonstration thermonuclear reactor-tokamak (T-20)

International Nuclear Information System (INIS)

Gusev, V.M.; Guseva, M.I.; Gervids, V.I.; Kogan, V.I.; Martynenko, Yu.V.; Mirnov, S.V.

A model is given for plasma interaction with the wall and the introduction of contaminants. The model was characterized by two kinds of uncertainty. First, the uncertain behavior of the contaminants, and second, the uncertainty of boundary conditions. Some of the conclusions from the study are described

First-principles calculations on double-walled inorganic nanotubes with hexagonal chiralities

Energy Technology Data Exchange (ETDEWEB)

Zhukovskii, Yuri F [Institute of Solid State Physics, University of Latvia, 8 Kengaraga Str., LV-1063, Riga (Latvia); Evarestov, Robert A; Bandura, Andrei V; Losev, Maxim V, E-mail: quantzh@latnet.lv [Department of Quantum Chemistry, St. Petersburg State University, 26 Universitetsky Ave., 198504, Petrodvorets (Russian Federation)

2011-06-23

The two sets of commensurate double-walled boron nitride and titania hexagonally-structured nanotubes (DW BN and TiO{sub 2} NTs) possessing either armchair- or zigzag-type chiralities have been considered, i.e., (n{sub 1},n{sub 1})-(n{sub 2},n{sub 2}) or (n{sub 1},0)-(n{sub 2},0), respectively. For symmetry analysis of these nanotubes, the line symmetry groups for one-periodic (1D) nanostructures with rotohelical symmetry have been applied. To analyze the structural and electronic properties of hexagonal DW NTs, a series of large-scale ab initio DFT-LCAO calculations have been performed using the hybrid Hartree-Fock/Kohn-Sham exchange-correlation functional PBE0 (as implemented in CRYSTAL-09 code). To establish the optimal inter-shell distances within DW NTs corresponding to the minima of calculated total energy, the chiral indices n{sub 1} and n{sub 2} of the constituent single-walled (SW) nanotubes have been successively varied.

Self-sustaining thin films as a means of reducing first wall erosion and plasma impurity influx

International Nuclear Information System (INIS)

Krauss, A.R.; Gruen, D.M.

1982-01-01

Neutral impurities ejected from Tokamak wall and limiter surfaces may travel several cm before being ionized very quickly upon entering the plasma edge. The influence of the unipolar sheath potential is exerted only within a very short distance of the surface and has no effect on neutral impurity atoms within a very short distance of the surface and has no effect on neutral impurity atoms which are subsequently ionized by charge-exchange collisions or electron impact ionization. However, secondary ions emanating from the limiter surfaces with kinetic energies less than the sheath potential will have essentially zero probability of traveling more than a few Debye lengths before being redeposited. Similarly, secondary ions originating at the first wall are redeposited as a result of the deflection produced by the magnetic field. Impurity influx resulting from sputtering would therefore be substantially reduced for surfaces which produce a very high ion/neutral ratio when sputtered. It has been previously shown that the high secondary ion yield associated with the alkali metal potassium does not apply to the bulk metal but pertains to ionic compounds and thin (mono-layer) films. Two processes are discussed as a means of producing these films in a self-sustaining manner compatible with the fusion reactor environment. (orig.)

Development of conductively cooled first wall armor and actively cooled divertor structure for ITER/FER

International Nuclear Information System (INIS)

Ioki, K.; Yamada, M.; Sakata, S.; Okada, K.; Toyoda, M.; Shimizu, K.; Tsujimura, S.; Iimura, M.; Akiba, M.; Araki, M.; Seki, M.

1991-01-01

Based on the design requirements for the plasma facing components in ITER/FER, we have performed design studies on the conductively cooled first wall armor and the divertor plate with sliding supports. The full-scale armor tiles were fabricated for heat load tests, and good thermal performances were obtained in heat load tests of 0.2-0.4 MW/m 2 . It is shown by the thermomechanical analysis on the divertor plate that thermal stresses and bending deformation are reduced significantly by using the sliding supports. The divertor test module with the sliding supports has been fabricated to investigate its fabricability and to verify the functions of the sliding supports during a high heat load of about 10 MW/m 2 . (orig.)

A Study of Aerodynamics in Kevlar-Wall Test Sections

OpenAIRE

Brown, Kenneth Alexander

2014-01-01

This study is undertaken to characterize the aerodynamic behavior of Kevlar-wall test sections and specifically those containing two-dimensional, lifting models. The performance of the Kevlar-wall test section can be evaluated against the standard of the hard-wall test section, which in the case of the Stability Wind Tunnel (SWT) at Virginia Tech can be alternately installed or replaced by the Kevlar-wall test section. As a first step towards the evaluation of the Kevlar-wall test section aer...

Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

International Nuclear Information System (INIS)

Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

1987-01-01

The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

Research on the wetted first wall concept for future laser fusion reactors. Final report No. 1, October 1, 1974--January 31, 1976

International Nuclear Information System (INIS)

Hoffman, M.A.; Munir, Z.A.

1976-01-01

Research is in progress to determine the feasibility of the wetted first wall concept for a future laser fusion reactor. The basic idea involves the use of a thin coating of lithium on the inner wall of the laser fusion containment vessel to protect it from the micro-explosion blast debris. This report contains a review of the available information on contact angles and wettability of alkali metals on various metal substrates as well as a review of literature on thin falling liquid films. A proposed experiment to measure the contact angles of lithium on stainless steel and niobium is described. The requirements for a second experiment to measure certain key characteristics of thin falling films are also included

Development of fatigue life criteria for experimental fusion reactor first-wall structures

International Nuclear Information System (INIS)

Nickell, R.E.; Esztegar, E.P.

1980-01-01

An approach to the rational design of fusion reactor first-wall structures against fatigue crack growth is proposed. The approach is motivated by microstructural observations of fatigue crack growth enhancement in uniruniradiated materials due to volumetric damage ahead of a propagating crack. Examples are cited that illustrate the effect of mean stress on void nucleation and coalescence, which represent the dominant form of volumetric damage at low temperature, and of grain boundary sliding and creep cavitation, which are the dominant volumetric damage mechanisms at high temperature. The analogy is then drawn between these forms of fatigue crack growth enhancement and those promoted by irradiation exposure in the fusion reactor environment, such as helium embrittlement and atomic displacement. An enhanced strain range is suggested as a macroscopic measure of the reduction in fatigue life due to the higher fatigue crack growth rates. The enhanced strain range permits a separation of volumetric and cyclic effects, and assists in the assignment of rational design factors to each effect. A series of experiments are outlined which should provide the numerical values of the parameters for the enhanced strain range. (orig.)

Aging near the wall in colloidal glasses

Science.gov (United States)

Cao, Cong; Huang, Xinru; Weeks, Eric

In a colloidal glass system, particles move slower as sample ages. In addition, their motions may be affected by their local structure, and this structure will be different near a wall. We examine how the aging process near a wall differs from that in the bulk of the sample. In particular, we use a confocal microscope to observe 3D motion in a bidisperse colloidal glass sample. We find that flat walls induce the particles to organize into layers. The aging process behaves differently near the boundary, especially within the first three layers. Particle motion near the wall is noticeably slower but also changes less dramatically with age. We compare and contrast aging seen in samples with flat and rough walls.

Dynamics of strings between walls

International Nuclear Information System (INIS)

Eto, Minoru; Fujimori, Toshiaki; Nagashima, Takayuki; Nitta, Muneto; Ohashi, Keisuke; Sakai, Norisuke

2009-01-01

Configurations of vortex strings stretched between or ending on domain walls were previously found to be 1/4 Bogomol'nyi-Prasad-Sommerfield (BPS) states in N=2 supersymmetric gauge theories in 3+1 dimensions. Among zero modes of string positions, the center of mass of strings in each region between two adjacent domain walls is shown to be non-normalizable whereas the rests are normalizable. We study dynamics of vortex strings stretched between separated domain walls by using two methods, the moduli space (geodesic) approximation of full 1/4 BPS states and the charged particle approximation for string end points in the wall effective action. In the first method we explicitly obtain the effective Lagrangian in the strong coupling limit, which is written in terms of hypergeometric functions, and find the 90 deg. scattering for head-on collision. In the second method the domain wall effective action is assumed to be U(1) N gauge theory, and we find a good agreement between two methods for well-separated strings.

The JET real-time plasma-wall load monitoring system

International Nuclear Information System (INIS)

Valcárcel, D.F.; Alves, D.; Card, P.; Carvalho, B.B.; Devaux, S.; Felton, R.; Goodyear, A.; Lomas, P.J.; Maviglia, F.; McCullen, P.; Reux, C.; Rimini, F.; Stephen, A.; Zabeo, L.

2014-01-01

Highlights: • The paper describes the JET real-time system monitoring the first-wall plasma loads. • It presents the motivation, physics basis, design and implementation of the system. • It also presents the integration in the JET CODAS. • Operational results are presented. - Abstract: In the past, the Joint European Torus (JET) has operated with a first-wall composed of Carbon Fibre Composite (CFC) tiles. The thermal properties of the wall were monitored in real-time during plasma operations by the WALLS system. This software routinely performed model-based thermal calculations of the divertor and Inner Wall Guard Limiter (IWGL) tiles calculating bulk temperatures and strike-point positions as well as raising alarms when these were beyond operational limits. Operation with the new ITER-like wall presents a whole new set of challenges regarding machine protection. One example relates to the new beryllium limiter tiles with a melting point of 1278 °C, which can be achieved during a plasma discharge well before the bulk temperature rises to this value. This requires new and accurate power deposition and thermal diffusion models. New systems were deployed for safe operation with the new wall: the Real-time Protection Sequencer (RTPS) and the Vessel Thermal Map (VTM). The former allows for a coordinated stop of the pulse and the latter uses the surface temperature map, measured by infra-red (IR) cameras, to raise alarms in case of hot-spots. Integration of WALLS with these systems is required as RTPS responds to raised alarms and VTM, the primary protection system for the ITER-like wall, can use WALLS as a vessel temperature provider. This paper presents the engineering design, implementation and results of WALLS towards D-T operation, where it will act as a primary protection system when the IR cameras are blinded by the fusion reaction neutrons. The first operational results, with emphasis on its performance, are also presented

The JET real-time plasma-wall load monitoring system

Energy Technology Data Exchange (ETDEWEB)

Valcárcel, D.F., E-mail: daniel.valcarcel@ipfn.ist.utl.pt [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, P-1049-001 Lisboa (Portugal); Alves, D. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, P-1049-001 Lisboa (Portugal); Card, P. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Carvalho, B.B. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, P-1049-001 Lisboa (Portugal); Devaux, S. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Felton, R.; Goodyear, A.; Lomas, P.J. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Univ. di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); McCullen, P. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Reux, C. [Ecole Polytechnique, LPP, CNRS UMR 7648, 91128 Palaiseau (France); Rimini, F.; Stephen, A. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Zabeo, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St., Paul Lez Durance (France); and others

2014-03-15

Highlights: • The paper describes the JET real-time system monitoring the first-wall plasma loads. • It presents the motivation, physics basis, design and implementation of the system. • It also presents the integration in the JET CODAS. • Operational results are presented. - Abstract: In the past, the Joint European Torus (JET) has operated with a first-wall composed of Carbon Fibre Composite (CFC) tiles. The thermal properties of the wall were monitored in real-time during plasma operations by the WALLS system. This software routinely performed model-based thermal calculations of the divertor and Inner Wall Guard Limiter (IWGL) tiles calculating bulk temperatures and strike-point positions as well as raising alarms when these were beyond operational limits. Operation with the new ITER-like wall presents a whole new set of challenges regarding machine protection. One example relates to the new beryllium limiter tiles with a melting point of 1278 °C, which can be achieved during a plasma discharge well before the bulk temperature rises to this value. This requires new and accurate power deposition and thermal diffusion models. New systems were deployed for safe operation with the new wall: the Real-time Protection Sequencer (RTPS) and the Vessel Thermal Map (VTM). The former allows for a coordinated stop of the pulse and the latter uses the surface temperature map, measured by infra-red (IR) cameras, to raise alarms in case of hot-spots. Integration of WALLS with these systems is required as RTPS responds to raised alarms and VTM, the primary protection system for the ITER-like wall, can use WALLS as a vessel temperature provider. This paper presents the engineering design, implementation and results of WALLS towards D-T operation, where it will act as a primary protection system when the IR cameras are blinded by the fusion reaction neutrons. The first operational results, with emphasis on its performance, are also presented.

Degradation processes and the methods of securing wall crests

Directory of Open Access Journals (Sweden)

Maciej Trochonowicz

2017-12-01

Full Text Available The protection of historical ruins requires solution of doctrinal and technical problems. Technical problems concern above all preservation of walls, which are exposed to the influence of atmospheric factors. The problem that needs to be solved in any historic ruin is securing of wall crests. Form of protection of the wall crests depends on many factors, mainly technical features of the wall and architectural and conservatory vision. The following article presents three aspects important for protection of wall crests. Firstly, analysis of features of the wall as a structure, secondly the characteristics of destructive agents, thirdly forms of protection of wall crests. In the summary of the following article, advantages and disadvantages of each method of preservation of the wall crests were presented.

Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

International Nuclear Information System (INIS)

Banetta, S.; Zacchia, F.; Lorenzetto, P.; Bobin-Vastra, I.; Boireau, B.; Cottin, A.; Mitteau, R.; Eaton, R.; Raffray, R.

2014-01-01

This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m 2 ) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles. The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares. In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time

Observations on resistive wall modes

International Nuclear Information System (INIS)

Gerwin, R.A.; Finn, J.M.

1996-01-01

Several results on resistive wall modes and their application to tokamaks are presented. First, it is observed that in the presence of collisional parallel dynamics there is an exact cancellation to lowest order of the dissipative and sound wave effects for an ideal Ohm's law. This is easily traced to the fact that the parallel dynamics occurs along the perturbed magnetic field lines for such electromagnetic modes. Such a cancellation does not occur in the resistive layer of a tearing-like mode. The relevance to models for resistive wall modes using an electrostatic Hammett-Perkins type operator to model Landau damping will be discussed. Second, we observe that with an ideal Ohm's law, resistive wall modes can be destabilized by rotation in that part of parameter space in which the ideal MHD modes are stable with the wall at infinity. This effect can easily be explained by interpreting the resistive wall instability in terms of mode coupling between the backward stable MHD mode and a stable mode locked into the wall. Such an effect can occur for very small rotation for tearing-resistive wall modes in which inertia dominates viscosity in the layer, but the mode is stabilized by further rotation. For modes for which viscosity dominates in the layer, rotation is purely stabilizing. For both tearing models, a somewhat higher rotation frequency gives stability essentially whenever the tearing mode is stable with a perfectly conducting wall. These tearing/resistive wall results axe also simply explained in terms of mode coupling. It has been shown that resonant external ideal modes can be stabilized in the presence of resistive wall and resistive plasma with rotation of order the nominal tearing mode growth rate. We show that these modes behave as resistive wall tearing modes in the sense above. This strengthens the suggestion that rotational stabilization of the external kink with a resistive wall is due to the presence of resistive layers, even for ideal modes

Simulation of surface cracks measurement in first walls by laser spot array thermography

Energy Technology Data Exchange (ETDEWEB)

Pei, Cuixiang; Qiu, Jinxin; Liu, Haocheng; Chen, Zhenmao, E-mail: chenzm@mail.xjtu.edu.cn

2016-11-01

The inspection of surface cracks in first walls (FW) is very important to ensure the safe operation of the fusion reactors. In this paper, a new laser excited thermography technique with using laser spot array source is proposed for the surface cracks imaging and evaluation in the FW with an intuitive and non-contact measurement method. Instead of imaging a crack by scanning a single laser spot and superimposing the local discontinuity images with the present laser excited thermography methods, it can inspect a relatively large area at one measurement. It does not only simplify the measurement system and data processing procedure, but also provide a faster measurement for FW. In order to investigate the feasibility of this method, a numerical code based on finite element method (FEM) is developed to simulate the heat flow and the effect of the crack geometry on the thermal wave fields. An imaging method based on the gradient of the thermal images is proposed for crack measurement with the laser spot array thermography method.

Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

Energy Technology Data Exchange (ETDEWEB)

Pérez, Germán, E-mail: german.perez.pichel@gmail.com; Mitteau, Raphaël; Eaton, Russell; Raffray, René

2015-12-15

Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

International Nuclear Information System (INIS)

Pérez, Germán; Mitteau, Raphaël; Eaton, Russell; Raffray, René

2015-01-01

Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

Energy Technology Data Exchange (ETDEWEB)

Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

2015-10-15

Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

Diaphragm walling for Sizewell B sets records

International Nuclear Information System (INIS)

Anon.

1988-01-01

The first phase of construction of the Sizewell-B nuclear reactor has been completed. This was the building of a diaphragm wall around the site. It is one of the largest and deepest diaphragm walls to be installed in Europe. The site can be pumped dry of groundwater and the foundations constructed in the dry. The specifications of the wall and its construction, using two Hydrofraise excavation rigs, are described. The excavated material is brought up as a slurry and the (bentonite) slurry is cleaned and desanded. Most of the wall has been formed using a plastic concrete but reinforced concrete has been used for some stretches. The diaphragm wall, which is 1258m long and 55m deep on average, was built in 19 weeks. (U.K.)

One-loop fluctuation-dissipation formula for bubble-wall velocity

International Nuclear Information System (INIS)

Arnold, P.

1993-01-01

The limiting bubble-wall velocity during a first-order electroweak phase transition is of interest in scenarios for electroweak baryogenesis. Khlebnikov has recently proposed an interesting method for computing this velocity based on the fluctuation-dissipation theorem. I demonstrate that at one-loop order this method is identical to simple, earlier techniques for computing the wall velocity based on computing the friction from particles reflecting off or transmitting through the wall in the ideal gas (''thin-wall'') limit

Automated image segmentation and registration of vessel wall MRI for quantitative assessment of carotid artery vessel wall dimensions and plaque composition

NARCIS (Netherlands)

Klooster, Ronald van 't

2014-01-01

The main goal of this thesis was to develop methods for automated segmentation, registration and classification of the carotid artery vessel wall and plaque components using multi-sequence MR vessel wall images to assess atherosclerosis. First, a general introduction into atherosclerosis and

Shock-tube study of fusion plasma-wall interactions

International Nuclear Information System (INIS)

Gross, R.A.; Tien, J.K.; Jensen, B.; Panayotou, N.F.; Feinberg, B.

1977-01-01

Theoretical and experimental studies have been made of phenomena which occur when a hot (T 1 approximately equal to 6 x 10 6 0 K), dense (n approximately equal to 10 16 cm -3 ), deuterium plasma containing a transverse magnetic field is brought into sudden contact with a cold metal wall. These studies are motivated by the need to understand plasma and metallurgical conditions at the first-wall of a fusion reactor. Experiments were carried out in the Columbia high energy electromagnetic shock tube. Computational simulation was used to investigate the detailed physics of the fusion plasma boundary layer which develops at the wall. The rate of energy transfer from the plasma to the wall was calculated and conditions under which surface melting occurs are estimated. Experimental measurements of plasma-wall heat transfer rates up to 3 x 10 5 watts/cm 2 were obtained and agreement with computed values are good. Fusion reactor first-wall materials have been exposed to 6.0 x 10 21 eV cm -2 (1,000 shots) of deuterium plasma bombardment. Scanning electron micrograph photographs show preferential erosion at grain boundaries, formation of deuterium surface blisters, and evidence of local surface melting. Some cracking is observed along grain boundaries, and a decrease in tensile ductiity is measured

Cost of quay walls including life cycle aspects

NARCIS (Netherlands)

De Gijt, J.G.; Vinks, R.

2011-01-01

Port authories and other organisations involved in designing and building of port infrastructure are at first glance interested in predicting adequatly the expected costs. This paper discusses the costs development of quay walls versus time. The basis for the costs development of quay walls is

Oxidation behaviour of silicon-free tungsten alloys for use as the first wall material

Science.gov (United States)

Koch, F.; Brinkmann, J.; Lindig, S.; Mishra, T. P.; Linsmeier, Ch

2011-12-01

The use of self-passivating tungsten alloys as armour material of the first wall of a fusion power reactor may be advantageous concerning safety issues. In earlier studies good performance of the system W-Cr-Si was demonstrated. Thin films of such alloys showed a strongly reduced oxidation rate compared to pure tungsten. However, the formation of brittle tungsten silicides may be disadvantageous for the powder metallurgical production of bulk W-Cr-Si alloys if a good workability is needed. This paper shows the results of screening tests to identify suitable silicon-free alloys with distinguished self-passivation and a potentially good workability. Of all the tested systems W-Cr-Ti alloys showed the most promising results. The oxidation rate was even lower than the one of W-Cr-Si alloys, the reduction factor was about four orders of magnitude compared to pure tungsten. This performance could be conserved even if the content of alloying elements was reduced.

Oxidation behaviour of silicon-free tungsten alloys for use as the first wall material

International Nuclear Information System (INIS)

Koch, F; Brinkmann, J; Lindig, S; Mishra, T P; Linsmeier, Ch

2011-01-01

The use of self-passivating tungsten alloys as armour material of the first wall of a fusion power reactor may be advantageous concerning safety issues. In earlier studies good performance of the system W-Cr-Si was demonstrated. Thin films of such alloys showed a strongly reduced oxidation rate compared to pure tungsten. However, the formation of brittle tungsten silicides may be disadvantageous for the powder metallurgical production of bulk W-Cr-Si alloys if a good workability is needed. This paper shows the results of screening tests to identify suitable silicon-free alloys with distinguished self-passivation and a potentially good workability. Of all the tested systems W-Cr-Ti alloys showed the most promising results. The oxidation rate was even lower than the one of W-Cr-Si alloys, the reduction factor was about four orders of magnitude compared to pure tungsten. This performance could be conserved even if the content of alloying elements was reduced.

Vaporized wall material/plasma interaction during plasma disruption

International Nuclear Information System (INIS)

Merrill, B.J.; Carroll, M.C.; Jardin, S.C.

1983-01-01

The purpose of this paper is to discuss a new plasma disruption model that has been developed for analyzing the consequences to the limiter/first wall structures. This model accounts for: nonequilibrium surface vaporization for the ablating structure, nonequilibrium ionization of and radiation emitted from the ablated material in the plasma, plasma particle and energy transport, and plasma electromagnetic field evolution during the disruption event. Calculations were performed for a 5 ms disruption on a stainless steel flat limiter as part of a D-shaped first wall. These results indicated that the effectiveness of the ablated wall material to shield the exposed structure is greater than predicted by earlier models, and that the rate of redeposition of the ablated wall material ions is very dramatic. Impurity transport along magnetic field lines, global plasma motion, and radiation transport in an optically thick plasma are important factors that require additional modeling. Experimental measurements are needed to verify these models

Coating requirements for an ICF dry-wall design

International Nuclear Information System (INIS)

Taylor, L.H.; Sucov, E.W.

1981-01-01

A new concept for protecting the first wall of an ICF reactor has been developed which relies heavily on a coating to protect the steel tubes which comprise the first wall. This coating must survive the pellet explosion, be ductile, and be compatible with the materials in the ICF pellet. Calculations indicate that tantalum is the best choice for the coating material and that tantalum coated steel tubes can handle fusion thermal powers of 3500 MW in a 10 m radius spherical chamber

Phenomenology of the domain walls in thin ferromagnetic films

International Nuclear Information System (INIS)

Adam, G.

1978-01-01

The basic concepts and the main theoretical methods developed in the study of the domain walls in thin ferromagnetic films are given in this review. First, an insight into the origins and the classification criteria of the conceptually different wall structures is obtained by elementary considerations which are mainly based on the experimentally available data. Then, the more subtle aspect of the wall models dimensionality in soft ferromagnetic films is discussed. Finally, the various theoretical calculation methods of the wall parameters are summarized. (author)

TiS2 and ZrS2 single- and double-wall nanotubes: first-principles study.

Science.gov (United States)

Bandura, Andrei V; Evarestov, Robert A

2014-02-15

Hybrid density functional theory has been applied for investigations of the electronic and atomic structure of bulk phases, nanolayers, and nanotubes based on titanium and zirconium disulfides. Calculations have been performed on the basis of the localized atomic functions by means of the CRYSTAL-2009 computer code. The full optimization of all atomic positions in the regarded systems has been made to study the atomic relaxation and to determine the most favorable structures. The different layered and isotropic bulk phases have been considered as the possible precursors of the nanotubes. Calculations on single-walled TiS2 and ZrS2 nanotubes confirmed that the nanotubes obtained by rolling up the hexagonal crystalline layers with octahedral 1T morphology are the most stable. The strain energy of TiS2 and ZrS2 nanotubes is small, does not depend on the tube chirality, and approximately obeys to D(-2) law (D is nanotube diameter) of the classical elasticity theory. It is greater than the strain energy of the similar TiO2 and ZrO2 nanotubes; however, the formation energy of the disulfide nanotubes is considerably less than the formation energy of the dioxide nanotubes. The distance and interaction energy between the single-wall components of the double-wall nanotubes is proved to be close to the distance and interaction energy between layers in the layered crystals. Analysis of the relaxed nanotube shape using radial coordinate of the metal atoms demonstrates a small but noticeable deviation from completely cylindrical cross-section of the external walls in the armchair-like double-wall nanotubes. Copyright © 2013 Wiley Periodicals, Inc.

Progress towards realization of a laser IFE solid wall chamber

International Nuclear Information System (INIS)

Raffray, A.R.; Blanchard, J.; Latkowski, J.; Najmabadi, F.; Renk, T.; Sethian, J.; Sharafat, S.; Snead, L.

2006-01-01

The high average power laser (HAPL) program aims at developing laser inertial fusion energy (Laser IFE) based on lasers, direct drive targets and a solid wall chamber. The preferred first wall configuration is based on tungsten and ferritic steel as armor and structural materials, respectively. A key concern is the survival of the first wall under the X-ray and ion energy deposition from the fusion micro-explosion. The HAPL design and R and D effort in the chamber and material area is focused toward understanding and resolving the key armor survival issues. This includes modeling and experimental testing of the armor thermo-mechanical behavior in facilities utilizing ion, X-rays and laser sources to simulate IFE conditions. Helium management is addressed by conducting implantation experiments along with modeling of He behavior in tungsten. This paper summarizes the HAPL chamber activities. The first wall/armor configuration and design analysis are described, key chamber issues are discussed, and the R and D to address them is highlighted

An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

Energy Technology Data Exchange (ETDEWEB)

Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

2006-07-05

An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperture of 700C. We have identified critical issues for the concept, some of which inlude the first wall design, the assessment of MHD effectrs with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time, we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

Benchmarking LES with wall-functions and RANS for fatigue problems in thermal–hydraulics systems

Energy Technology Data Exchange (ETDEWEB)

Tunstall, R., E-mail: ryan.tunstall@manchester.ac.uk [School of MACE, The University of Manchester, Manchester M13 9PL (United Kingdom); Laurence, D.; Prosser, R. [School of MACE, The University of Manchester, Manchester M13 9PL (United Kingdom); Skillen, A. [Scientific Computing Department, STFC Daresbury Laboratory, Warrington WA4 4AD (United Kingdom)

2016-11-15

Highlights: • We benchmark LES with blended wall-functions and low-Re RANS for a pipe bend and T-Junction. • Blended wall-laws allow the first cell from the wall to be placed anywhere in the boundary layer. • In both cases LES predictions improve as the first cell wall spacing is reduced. • Near-wall temperature fluctuations in the T-Junction are overpredicted by wall-modelled LES. • The EBRSM outperforms other RANS models for the pipe bend. - Abstract: In assessing whether nuclear plant components such as T-Junctions are likely to suffer thermal fatigue problems in service, CFD techniques need to provide accurate predictions for wall temperature fluctuations. Though it has been established that this is within the capabilities of wall-resolved LES, its high computational cost has prevented widespread usage in industry. In the present paper the suitability of LES with blended wall-functions, that allow the first cell to be placed in any part of the boundary layer, is assessed. Numerical results for the flows through a 90° pipe bend and a T-Junction are compared against experimental data. Both test cases contain areas where equilibrium laws are violated in practice. It is shown that reducing the first cell wall spacing improves agreement with experimental data by limiting the extent from the wall in which the solution is constrained to an equilibrium law. The LES with wall-function approach consistently overpredicts the near-wall temperature fluctuations in the T-Junction, suggesting that it can be considered as a conservative approach. We also benchmark a range of low-Re RANS models. EBRSM predictions for the 90° pipe bend are in significantly better agreement with experimental data than those from the other models. There are discrepancies from all RANS models in the case of the T-Junction.

Condensation on a cooled plane upright wall

International Nuclear Information System (INIS)

Fortier, Andre.

1975-01-01

The vapor condensation along a cooled upright plane wall was studied. The theoretical and experimental results obtained in the simple case, give the essential characteristics of the phenomenon of condensation along a cold wall that keeps the vapor apart from the coolant inside a surface condenser. The phenomenon presents two different appearances according as the wall is wetted or not by the liquid. In the first case a continuous liquid film runs down the wall and a conventional Nusselt calculation gives the film thickness and the heat exchange coefficient between a pure saturated vapor and the cold wall. The calculation is developed in detail and the effect of a vapor flow along the film is discussed as well as that of the presence of a noncondensable gas inside the vapor. In the second case, separated liquid drops are formed on the wall, the phenomenon is called ''dropwise condensation'' and the heat exchange coefficients obtained are much higher than with film condensation. The theoretical aspects of the problem are discussed with some experimental results [fr

Improving Station Performance by Building Isolation Walls in the Tunnel

Science.gov (United States)

Jia, Yan; Horn, Nikolaus; Leohardt, Roman

2014-05-01

Conrad Observatory is situated far away from roads and industrial areas on the Trafelberg in Lower Austria. At the end of the seismic tunnel, the main seismic instrument of the Observatory with a station code CONA is located. This station is one of the most important seismic stations in the Austrian Seismic Network (network code OE). The seismic observatory consists of a 145m long gallery and an underground laboratory building with several working areas. About 25 meters away from the station CONA, six temporary seismic stations were implemented for research purposes. Two of them were installed with the same equipment as CONA, while the remaining four stations were set up with digitizers having lower noise and higher resolution (Q330HR) and sensors with the same type (STS-2). In order to prevent possible disturbances by air pressure and temperature fluctuation, three walls were built inside of the tunnel. The first wall is located ca 63 meters from the tunnel entrance, while a set of double walls with a distance of 1.5 meters is placed about 53 meters from the first isolation wall but between the station CONA and the six temporary stations. To assess impact of the isolation walls on noise reduction and detection performance, investigations are conducted in two steps. The first study is carried out by comparing the noise level and detection performance between the station CONA behind the double walls and the stations in front of the double walls for verifying the noise isolation by the double walls. To evaluate the effect of the single wall, station noise level and detection performance were studied by comparing the results before and after the installation of the wall. Results and discussions will be presented. Additional experiment is conducted by filling insulation material inside of the aluminium boxes of the sensors (above and around the sensors). This should help us to determine an optimal insulation of the sensors with respect to pressure and temperature

Wall shear stress measurement of near-wall flow over inclined and curved boundaries by stereo interfacial particle image velocimetry

International Nuclear Information System (INIS)

Nguyen, Thien Duy; Wells, John Craig; Nguyen, Chuong Vinh

2010-01-01

In investigations of laminar or turbulent flows, wall shear is often important. Nevertheless, conventional particle image velocimetry (PIV) is difficult in near-wall regions. A near-wall measurement technique, named interfacial PIV (IPIV) [Nguyen, C., Nguyen, T., Wells, J., Nakayama, A., 2008. Proposals for PIV of near-wall flow over curved boundaries. In: Proceedings of 14th International Symposium on Applications of Laser Technique to Fluid Mechanics], handles curved boundaries by means of conformal transformation, directly measures the wall gradient, and yields the near-wall tangential velocity profile at one-pixel resolution. In this paper, we show the feasibility of extending IPIV to measure wall gradients by stereo reconstruction. First, we perform a test on synthetic images generated from a direct numerical simulation (DNS) snapshot of turbulent flow over sinusoidal bed. Comparative assessment of wall gradients derived by IPIV, stereo-IPIV and particle image distortion (PID) [Huang, H.T., Fiedler, H.E., Wang, J.J., 1993. Limitation and improvement of PIV. Experiments in Fluids 15(4), 263-273] is evaluated with DNS data. Also, the sensitivity of IPIV and stereo-IPIV results to the uncertainty of identified wall position is examined. As a practical application of IPIV and stereo-IPIV to experimental images, results from turbulent open channel flow over a backward-facing step are discussed in detail.

Dynamical evolution of domain walls in an expanding universe

Science.gov (United States)

Press, William H.; Ryden, Barbara S.; Spergel, David N.

1989-01-01

Whenever the potential of a scalar field has two or more separated, degenerate minima, domain walls form as the universe cools. The evolution of the resulting network of domain walls is calculated for the case of two potential minima in two and three dimensions, including wall annihilation, crossing, and reconnection effects. The nature of the evolution is found to be largely independent of the rate at which the universe expands. Wall annihilation and reconnection occur almost as fast as causality allows, so that the horizon volume is 'swept clean' and contains, at any time, only about one, fairly smooth, wall. Quantitative statistics are given. The total area of wall per volume decreases as the first power of time. The relative slowness of the decrease and the smoothness of the wall on the horizon scale make it impossible for walls to both generate large-scale structure and be consistent with quadrupole microwave background anisotropy limits.

KETERASINGAN DALAM FILM WALL-E

Directory of Open Access Journals (Sweden)

Rahmadya Putra Nugraha

2017-05-01

Full Text Available Modern society nowadays technological advances at first create efficiency in human life. Further development of the technology thus drown human in a routine and automation of work created. The State is to be one of the causes of man separated from fellow or the outside world and eventually experiencing alienation. The movie as a mass media function to obtain the movie and entertainment can be informative or educative function is contained, even persuasive. The purpose of this research was conducted to find out the alienation in the movie Wall E. The concepts used to analyze the movie Wall E this is communication, movie, and alienation. The concept of alienation of human alienation from covering its own products of human alienation from its activities, the human alienation from nature of his humanity and human alienation from each other. Paradigm used is a critical paradigm with type a descriptive research with qualitative approach. The method used is the analysis of semiotics Roland Barthes to interpretation the scope of social alienation and fellow humans in the movie.This writing research results found that alienation of humans with other humans influenced the development of the technology and how the human it self represented of technology, not from our fellow human beings. Masyarakat modern saat ini kemajuan teknologi pada awalnya membuat efisiensi dalam kehidupan manusia. Perkembangan selanjutnya teknologi justru menenggelamkan manusia dalam suatu rutinitas dan otomatisasi kerja yang diciptakan. Keadaan itulah yang menjadi salah satu penyebab manusia terpisah dari sesama atau dunia luar dan akhirnya mengalami keterasingan. Film sebagai media massa berfungsi untuk memperoleh hiburan dan dalam film dapat terkandung fungsi informatif maupun edukatif, bahkan persuasif. Tujuan Penelitian ini dilakukan untuk mengetahui Keterasingan dalam film Wall E. Konsep-konsep yang digunakan untuk menganalisis film Wall E ini adalah komunikasi, film, dan

Impact of wall potential on the fluid-wall interaction in a cylindrical capillary and a generalized Kelvin equation

International Nuclear Information System (INIS)

Jakubov, T.S.; Mainwaring, D.E.

2006-01-01

In the present work a generalized Kelvin equation for a fluid confined in thick-walled cylindrical capillary is developed. This has been accomplished by including the potential energy function for interaction between a solid wall of a capillary and a confined fluid into the Kelvin equation. Using the Lennard-Jones 12-6 potential, an explicit form of the potential energy functions as expressed by hypergeometrical functions have been derived-firstly, for the interaction between a solid wall and a test atom placed at an arbitrary point in a long open-end capillary, and thereafter for the body-body interaction between the solid wall and a confined Lennard-Jones fluid. Further, this generalized Kelvin equation has been applied to detailed description hysteresis phenomena in such capillaries. All numerical calculations have been carried out for the model argon-graphite system at 90 K

Effect of design geometry of the demo first wall on the plasma heat load

Directory of Open Access Journals (Sweden)

Yu. Igitkhanov

2016-12-01

Full Text Available In this work we analyse the effect of W armour surface shaping on the heat load on the W/EUROFER DEMO sandwich type first wall blanket module with the water coolant. The armour wetted area is varied by changing the inclination and height of the «roof» type armor surface. The deleterious effect of leading edge at the tiles corner caused by misalignment is replaced in current design by rounded corners. Analysis has been carried out by means of the MEMOS code to assess the influence of the thickness of the layers and effect of the magnetic field inclination. Calculations show the evolution of the maximum temperatures in the tungsten, EUROFER, Cu allow and the stainless-steel water tube for different level of surface inclination (chamfering and in the case of rounded corners used in the current design. It is shown that the blanket module materials remain within a proper temperature range only at shallow incident angle if the width of EUROFER is reduced at list twice compare with the reference case.

Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak

Science.gov (United States)

Labate, C.; Di Gironimo, G.; Renno, F.

2015-09-01

Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.

Sunspot Light Walls Suppressed by Nearby Brightenings

Energy Technology Data Exchange (ETDEWEB)

Yang, Shuhong; Zhang, Jun; Hou, Yijun; Li, Xiaohong [CAS Key Laboratory of Solar Activity, National Astronomical Observatories, Chinese Academy of Sciences, Beijing 100012 (China); Erdélyi, Robertus [Solar Physics and Space Plasma Research Centre, School of Mathematics and Statistics, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Yan, Limei, E-mail: shuhongyang@nao.cas.cn [Key Laboratory of Earth and Planetary Physics, Institute of Geology and Geophysics, Chinese Academy of Sciences, Beijing 100029 (China)

2017-07-01

Light walls, as ensembles of oscillating bright structures rooted in sunspot light bridges, have not been well studied, although they are important for understanding sunspot properties. Using the Interface Region Imaging Spectrograph and Solar Dynamics Observatory observations, here we study the evolution of two oscillating light walls each within its own active region (AR). The emission of each light wall decays greatly after the appearance of adjacent brightenings. For the first light wall, rooted within AR 12565, the average height, amplitude, and oscillation period significantly decrease from 3.5 Mm, 1.7 Mm, and 8.5 minutes to 1.6 Mm, 0.4 Mm, and 3.0 minutes, respectively. For the second light wall, rooted within AR 12597, the mean height, amplitude, and oscillation period of the light wall decrease from 2.1 Mm, 0.5 Mm, and 3.0 minutes to 1.5 Mm, 0.2 Mm, and 2.1 minutes, respectively. Particularly, a part of the second light wall even becomes invisible after the influence of a nearby brightening. These results reveal that the light walls are suppressed by nearby brightenings. Considering the complex magnetic topology in light bridges, we conjecture that the fading of light walls may be caused by a drop in the magnetic pressure, where the flux is canceled by magnetic reconnection at the site of the nearby brightening. Another hypothesis is that the wall fading is due to the suppression of driver source ( p -mode oscillation), resulting from the nearby avalanche of downward particles along reconnected brightening loops.

Overview of impurity control and wall conditioning in NSTX

International Nuclear Information System (INIS)

Kugel, H.W.; Maingi, R.; Wampler, W.; Barry, R.E.; Bell, M.; Blanchard, W.; Gates, D.; Johnson, D.; Kaita, R.; Kaye, S.; Maqueda, R.; Menard, J.; Menon, M.M.; Mueller, D.; Ono, M.; Paul, S.; Peng, Y-K.M.; Raman, R.; Roquemore, A.; Skinner, C. H.; Sabbagh, S.; Stratton, B.; Stutman, D.; Wilson, J. R.; Zweben, S.

2000-01-01

The National Spherical Torus Experiment (NSTX) started plasma operations i n February 1999. In the first extended period of experiments, NSTX achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. As expected, discharge reproducibility and performance were strongly affected by wall conditions. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results

One-loop fluctuation-dissipation formula for bubble-wall velocity

International Nuclear Information System (INIS)

Arnold, P.

1993-01-01

The limiting bubble wall velocity during a first-order electroweak phase transition is of interest in scenarios for electroweak baryogenesis. Khlebnikov has recently proposed an interesting method for computing this velocity based on the fluctuation-dissipation theorem. It is demonstrated that at one-loop order this method is identical to simple, earlier techniques for computing the wall velocity based on computing the friction from particles reflecting off or transmitting through the wall in the ideal gas limit

Quantitative contrast-enhanced first-pass cardiac perfusion MRI at 3 tesla with accurate arterial input function and myocardial wall enhancement.

Science.gov (United States)

Breton, Elodie; Kim, Daniel; Chung, Sohae; Axel, Leon

2011-09-01

To develop, and validate in vivo, a robust quantitative first-pass perfusion cardiovascular MR (CMR) method with accurate arterial input function (AIF) and myocardial wall enhancement. A saturation-recovery (SR) pulse sequence was modified to sequentially acquire multiple slices after a single nonselective saturation pulse at 3 Tesla. In each heartbeat, an AIF image is acquired in the aortic root with a short time delay (TD) (50 ms), followed by the acquisition of myocardial images with longer TD values (∼150-400 ms). Longitudinal relaxation rates (R(1) = 1/T(1)) were calculated using an ideal saturation recovery equation based on the Bloch equation, and corresponding gadolinium contrast concentrations were calculated assuming fast water exchange condition. The proposed method was validated against a reference multi-point SR method by comparing their respective R(1) measurements in the blood and left ventricular myocardium, before and at multiple time-points following contrast injections, in 7 volunteers. R(1) measurements with the proposed method and reference multi-point method were strongly correlated (r > 0.88, P < 10(-5)) and in good agreement (mean difference ±1.96 standard deviation 0.131 ± 0.317/0.018 ± 0.140 s(-1) for blood/myocardium, respectively). The proposed quantitative first-pass perfusion CMR method measured accurate R(1) values for quantification of AIF and myocardial wall contrast agent concentrations in 3 cardiac short-axis slices, in a total acquisition time of 523 ms per heartbeat. Copyright © 2011 Wiley-Liss, Inc.

The first cut-off wall in the Indian Himalayas for the dam of the Dhauliganga hydroelectric project

Energy Technology Data Exchange (ETDEWEB)

Brunner, W.G. [Bauer Maschinen GmbH, Berlin (Germany)

2006-07-01

This paper provided details a Bauer cutter used to build a cut-off wall for the Dhaulinganga power plant project in the Himalayan mountains. The dam for the project was built as a 56 m high concrete-faced rockfill dam with a length of 270 m at the crown. A cut-off wall was constructed on the upstream side of the dam extending down from the dam's plinth to the bedrock level. A Bauer cutter was used to key the cut-off wall straight into the bedrock, which omitted the need for a grout curtain. The cut-off wall is 1 m thick and 70 m deep, with a total area of 8000 m{sup 2}. The wall was constructed as a series of primary and secondary panels. Excavation of the panels was carried out in single bites by the Bauer DHG hydraulic diaphragm wall grabs, supported a box chisel, cross chisel and a Bauer BC 40 rock cutter. Trench stability was provided by bentonite slurry. The closing forces were activated by a cylinder which was installed vertically inside the base body. The Bauer cutter continuously removed soil and rock from the bottom of the trench for mixing with the bentonite slurry. The slurry was then pumped through a ring main of hose pipes to a desanding plant where it was cleaned and returned to the trench. Advantages offered by using the cutter included a consistently high output, an extremely high degree of verticality, watertight joints, and the ability to cut through hard boulders. Use of the cutter at the Dhaulinganga site showed that the project could not be carried out successfully without the use of the cutter, which was used whenever grab and chisel methods were unable to achieve satisfactory rates of penetration. Deployment of the cutter was essential to key the cut-off wall into the underlying bedrock. It was concluded that the Dhualinganga project will provide a model for future power generation projects in the Indian Himalayas. 11 figs.

Characteristics of wall pressure over wall with permeable coating

Energy Technology Data Exchange (ETDEWEB)

Song, Woo Seog; Shin, Seungyeol; Lee, Seungbae [Inha Univ., Incheon (Korea, Republic of)

2012-11-15

Fluctuating wall pressures were measured using an array of 16 piezoelectric transducers beneath a turbulent boundary layer. The coating used in this experiment was an open cell, urethane type foam with a porosity of approximately 50 ppi. The ultimate objective of the coating is to provide a mechanical filter to reduce the wall pressure fluctuations. The ultimate objective of the coating is to provide a mechanical filter to reduce the wall pressure fluctuations. The boundary layer on the flat plate was measured by using a hot wire probe, and the CPM method was used to determine the skin friction coefficient. The wall pressure autospectra and streamwise wavenumber frequency spectra were compared to assess the attenuation of the wall pressure field by the coating. The coating is shown to attenuate the convective wall pressure energy. However, the relatively rough surface of the coating in this investigation resulted in a higher mean wall shear stress, thicker boundary layer, and higher low frequency wall pressure spectral levels compared to a smooth wall.

Light shining through walls

Energy Technology Data Exchange (ETDEWEB)

Redondo, Javier [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Max-Planck-Institut fuer Physik, Muenchen (Germany); Ringwald, Andreas [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

2010-11-15

Shining light through walls? At first glance this sounds crazy. However, very feeble gravitational and electroweak effects allow for this exotic possibility. Unfortunately, with present and near future technologies the opportunity to observe light shining through walls via these effects is completely out of question. Nevertheless there are quite a number of experimental collaborations around the globe involved in this quest. Why are they doing it? Are there additional ways of sending photons through opaque matter? Indeed, various extensions of the standard model of particle physics predict the existence of new particles called WISPs - extremely weakly interacting slim particles. Photons can convert into these hypothetical particles, which have no problems to penetrate very dense materials, and these can reconvert into photons after their passage - as if light was effectively traversing walls. We review this exciting field of research, describing the most important WISPs, the present and future experiments, the indirect hints from astrophysics and cosmology pointing to the existence of WISPs, and finally outlining the consequences that the discovery of WISPs would have. (orig.)

Light shining through walls

International Nuclear Information System (INIS)

Redondo, Javier; Ringwald, Andreas

2010-11-01

Shining light through walls? At first glance this sounds crazy. However, very feeble gravitational and electroweak effects allow for this exotic possibility. Unfortunately, with present and near future technologies the opportunity to observe light shining through walls via these effects is completely out of question. Nevertheless there are quite a number of experimental collaborations around the globe involved in this quest. Why are they doing it? Are there additional ways of sending photons through opaque matter? Indeed, various extensions of the standard model of particle physics predict the existence of new particles called WISPs - extremely weakly interacting slim particles. Photons can convert into these hypothetical particles, which have no problems to penetrate very dense materials, and these can reconvert into photons after their passage - as if light was effectively traversing walls. We review this exciting field of research, describing the most important WISPs, the present and future experiments, the indirect hints from astrophysics and cosmology pointing to the existence of WISPs, and finally outlining the consequences that the discovery of WISPs would have. (orig.)

An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

Energy Technology Data Exchange (ETDEWEB)

Wong, C.P.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)]. E-mail: wongc@fusion.gat.com; Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Sawan, M. [University of Wisconsin, Madison, WI (United States); Dagher, M. [University of California, Los Angeles, CA (United States); Smolentsev, S. [University of California, Los Angeles, CA (United States); Merrill, B. [INEEL, Idaho Falls, ID (United States); Youssef, M. [University of California, Los Angeles, CA (United States); Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sze, D.K. [University of California, San Diego, CA (United States); Morley, N.B. [University of California, Los Angeles, CA (United States); Sharafat, S. [University of California, Los Angeles, CA (United States); Calderoni, P. [University of California, Los Angeles, CA (United States); Sviatoslavsky, G. [University of Wisconsin, Madison, WI (United States); Kurtz, R. [Pacific Northwest Laboratory, Richland, WA (United States); Fogarty, P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Zinkle, S. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Abdou, M. [University of California, Los Angeles, CA (United States)

2006-02-15

An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC{sub f}/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 deg. C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R and D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

Force-displacement response of unreinforced masonry walls for seismic design

International Nuclear Information System (INIS)

Petry, S.

2015-01-01

This thesis submitted to the Swiss Federal Institute of Technology EPFL contributes to the improvement of the design and assessment methods for unreinforced masonry (URM) wall structures built with modern hollow core clay bricks. First, an experimental campaign on the lateral nonlinear in-plane response of URM walls is presented; secondly, an existing dataset on URM walls is extended and reanalysed. A newly developed mechanical model which describes the full force-displacement response of URM walls is described. Two series of URM walls tested under lateral in-plane loading are presented. Throughout the quasi-cyclic tests of all URM walls, the deformations were recorded using a digital photogrammetric measurement system which tracked the displacement field of the walls. Based on these findings, a new mechanical model is proposed which describes the nonlinear force-displacement response of flexural dominated URM walls up to near collapse

Compositional change of some first wall materials by considering multiple step nuclear reaction

Energy Technology Data Exchange (ETDEWEB)

Noda, Tetsuji; Utsumi, Misako; Fujita, Mitsutane [National Research Inst. for Metals, Tsukuba, Ibaraki (Japan)

1997-03-01

The conceptual system for nuclear material design is considered and some trials on WWW server with functions of the easily accessible simulation of nuclear reactions are introduced. Moreover, as an example of the simulation on the system using nuclear data, transmutation calculation was made for candidate first wall materials such as 9Cr-2W steel, V-5Cr-5Ti and SiC in SUS316/Li{sub 2}O/H{sub 2}O(SUS), 9Cr-2WLi{sub 2}O/H{sub 2}O(RAF), V alloy/Li/Be(V), and SiC/Li{sub 2}ZrO{sub 3}/He(SiC) blanket/shield systems based on ITER design model. Neutron spectrum varies with different blanket/shield compositions. The flux of low energy neutrons decreases in order of V-SiC-RAF-SUS blanket/shield systems. Fair amounts of W depletion in 9Cr-2W steel and the increase of Cr content in V-5Cr-5Ti were predicted in SUS or RAF systems. Concentration change in W and Cr is estimated to be suppressed if Li coolant is used in place of water. Helium and hydrogen production are not strongly affected by the different blanket/shield compositions. (author)

INTEGRATED ENERGY EFFICIENT WINDOW-WALL SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Michael Arney, Ph.D.

2002-12-31

The building industry faces the challenge of reducing energy use while simultaneously improving construction methods and marketability. This paper describes the first phase of a project to address these concerns by designing an Integrated Window Wall System (IWWS) that can be commercialized. This work builds on previous research conducted during the 1990's by Lawrence Berkeley national Laboratories (LBNL). During this phase, the objective was to identify appropriate technologies, problems and issues and develop a number of design concepts. Four design concepts were developed into prototypes and preliminary energy analyses were conducted Three of these concepts (the foam wall, steel wall, and stiffened plate designs) showed particular potential for meeting the project objectives and will be continued into a second phase where one or two of the systems will be brought closer to commercialization.

Near wall combustion modeling in spark ignition engines. Part A: Flame–wall interaction

International Nuclear Information System (INIS)

Demesoukas, Sokratis; Caillol, Christian; Higelin, Pascal; Boiarciuc, Andrei; Floch, Alain

2015-01-01

Highlights: • A model for flame–wall interaction in addition to flame wrinkling by turbulence is proposed. • Two sparkplug positions and two lengths are used in a test engine for model validation. • Flame–wall interaction decreases the maximum values of cylinder pressure and heat release rates. • The impact of combustion chamber geometry is taken into account by the flame–wall interaction model. - Abstract: Research and design in the field of spark ignition engines seek to achieve high performance while conserving fuel economy and low pollutant emissions. For the evaluation of various engine configurations, numerical simulations are favored, since they are quick and less expensive than experiments. Various zero-dimensional combustion models are currently used. Both flame front reactions and post-flame processes contribute to the heat release rate. The first part of this study focuses on the role of the flame front on the heat release rate, by modeling the interaction of the flame front with the chamber wall. Post-flame reactions are dealt with in Part B of the study. The basic configurations of flame quenching in laminar flames are also applicable in turbulent flames, which is the case in spark ignition engines. A simplified geometric model of the combustion chamber was used to calculate the mean flame surface, the flame volume and the distribution of flame surface as a function of the distance from the wall. The flame–wall interaction took into account the geometry of the combustion chamber and of the flame, aerodynamic turbulence and the in-cylinder pressure and temperature conditions, through a phenomenological attenuation function of the wrinkling factor. A modified global wrinkling factor as a function of the mean surface distance distribution from the wall was calculated. The impact of flame–wall interaction was simulated for four configurations of the sparkplug position and length: centered and lateral position, and standard and projected

Reinforcement mechanism of multi-anchor wall with double wall facing

Science.gov (United States)

Suzuki, Kouta; Kobayashi, Makoto; Miura, Kinya; Konami, Takeharu; Hayashi, Taketo

2017-10-01

The reinforced soil wall has high seismic performance as generally known. However, the seismic behavior has not been clarified accurately yet, especially on multi-anchor wall with double wall facing. Indefinite behavior of reinforced soil wall during earthquake make us complicated in case with adopting to the abutment, because of arrangement of anchor plate as reinforcement often different according to the width of roads. In this study, a series of centrifuge model tests were carried out to investigate the reinforcement mechanism of multi anchor wall with double wall facing from the perspective of the vertical earth pressure. Several types of reinforce arrangement and rigid wall were applied in order to verify the arch function in the reinforced regions. The test results show unique behavior of vertical earth pressure, which was affected by arch action. All the vertical earth pressure placed behind facing panel, are larger than that of middle part between facing panel despite of friction between backfill and facing panel. Similar results were obtained in case using rigid wall. On the other hands, the vertical earth pressure, which were measured at the 3cm high from bottom of model container, shows larger than that of bottom. This results show the existence of arch action between double walls. In addition, it implies that the wall facing of such soil structure confined the backfill as pseudo wall, which is very reason that the multi anchor wall with double wall facing has high seismic performance.

Spin motive forces due to magnetic vortices and domain walls

NARCIS (Netherlands)

Lucassen, M.E.; Kruis, G.C.F.L.; Lavrijsen, R.; Swagten, H.J.M.; Koopmans, B.; Duine, R.A.

2011-01-01

We study spin motive forces, that is, spin-dependent forces and voltages induced by time-dependent magnetization textures, for moving magnetic vortices and domain walls. First, we consider the voltage generated by a one-dimensional field-driven domain wall. Next, we perform detailed calculations on

Contribution of domain wall networks to the CMB power spectrum

International Nuclear Information System (INIS)

Lazanu, A.; Martins, C.J.A.P.; Shellard, E.P.S.

2015-01-01

We use three domain wall simulations from the radiation era to the late-time dark energy domination era based on the PRS algorithm to calculate the energy–momentum tensor components of domain wall networks in an expanding universe. Unequal time correlators in the radiation, matter and cosmological constant epochs are calculated using the scaling regime of each of the simulations. The CMB power spectrum of a network of domain walls is determined. The first ever quantitative constraint for the domain wall surface tension is obtained using a Markov chain Monte Carlo method; an energy scale of domain walls of 0.93 MeV, which is close but below the Zel'dovich bound, is determined

Contribution of domain wall networks to the CMB power spectrum

Energy Technology Data Exchange (ETDEWEB)

Lazanu, A., E-mail: A.Lazanu@damtp.cam.ac.uk [Centre for Theoretical Cosmology, Department of Applied Mathematics and Theoretical Physics, Wilberforce Road, Cambridge CB3 0WA (United Kingdom); Martins, C.J.A.P., E-mail: Carlos.Martins@astro.up.pt [Centro de Astrofísica, Universidade do Porto, Rua das Estrelas, 4150-762 Porto (Portugal); Instituto de Astrofísica e Ciências do Espaço, CAUP, Rua das Estrelas, 4150-762 Porto (Portugal); Shellard, E.P.S., E-mail: E.P.S.Shellard@damtp.cam.ac.uk [Centre for Theoretical Cosmology, Department of Applied Mathematics and Theoretical Physics, Wilberforce Road, Cambridge CB3 0WA (United Kingdom)

2015-07-30

We use three domain wall simulations from the radiation era to the late-time dark energy domination era based on the PRS algorithm to calculate the energy–momentum tensor components of domain wall networks in an expanding universe. Unequal time correlators in the radiation, matter and cosmological constant epochs are calculated using the scaling regime of each of the simulations. The CMB power spectrum of a network of domain walls is determined. The first ever quantitative constraint for the domain wall surface tension is obtained using a Markov chain Monte Carlo method; an energy scale of domain walls of 0.93 MeV, which is close but below the Zel'dovich bound, is determined.

Contribution of domain wall networks to the CMB power spectrum

Directory of Open Access Journals (Sweden)

A. Lazanu

2015-07-01

Full Text Available We use three domain wall simulations from the radiation era to the late-time dark energy domination era based on the PRS algorithm to calculate the energy–momentum tensor components of domain wall networks in an expanding universe. Unequal time correlators in the radiation, matter and cosmological constant epochs are calculated using the scaling regime of each of the simulations. The CMB power spectrum of a network of domain walls is determined. The first ever quantitative constraint for the domain wall surface tension is obtained using a Markov chain Monte Carlo method; an energy scale of domain walls of 0.93 MeV, which is close but below the Zel'dovich bound, is determined.

Domain walls in single-chain magnets

Science.gov (United States)

Pianet, Vivien; Urdampilleta, Matias; Colin, Thierry; Clérac, Rodolphe; Coulon, Claude

2017-12-01

The topology and creation energy of domain walls in different magnetic chains (called Single-Chain Magnets or SCMs) are discussed. As these domain walls, that can be seen as "defects", are known to control both static and dynamic properties of these one-dimensional systems, their study and understanding are necessary first steps before a deeper discussion of the SCM properties at finite temperature. The starting point of the paper is the simple regular ferromagnetic chain for which the characteristics of the domain walls are well known. Then two cases will be discussed (i) the "mixed chains" in which isotropic and anisotropic classical spins alternate, and (ii) the so-called "canted chains" where two different easy axis directions are present. In particular, we show that "strictly narrow" domain walls no longer exist in these more complex cases, while a cascade of phase transitions is found for canted chains as the canting angle approaches 45∘. The consequence for thermodynamic properties is briefly discussed in the last part of the paper.

Initial phase wall conditioning in KSTAR

International Nuclear Information System (INIS)

Hong, Suk-Ho; Kim, Kwang-Pyo; Kim, Sungwoo; Lee, Dong-Su; Kim, Kyung-Min; Lee, Kun-Su; Kim, Jong-Su; Park, Jae-Min; Kim, Woong-Chae; Kim, Hak-Kun; Park, Kap-Rai; Yang, Hyung-Lyeol; Sun, Jong-Ho; Woo, Hyun-Jong; Lee, Sang-Yong; Lee, Sang-Hwa; Park, Eun-Kyung; Park, Sang-Joon; Kim, Sun-Ho; Wang, Sun-Jung

2011-01-01

The initial phase wall conditioning in KSTAR is depicted. The KSTAR wall conditioning procedure consists of vessel baking, glow discharge cleaning (GDC), ICRH wall conditioning (ICWC) and boronization (Bz). Vessel baking is performed for the initial vacuum conditioning in order to remove various kinds of impurities including H 2 O, carbon and oxygen and for the plasma operation. The total outgassing rates after vessel baking in three successive KSTAR campaigns are compared. GDC is regularly performed as a standard wall cleaning procedure. Another cleaning technique is ICWC, which is useful for inter-shot wall conditioning under a strong magnetic field. In order to optimize the operation time and removal efficiency of ICWC, a parameter scan is performed. Bz is a standard technique to remove oxygen impurity from a vacuum vessel. KSTAR has used carborane powder which is a non-toxic boron-containing material. The KSTAR Bz has been successfully performed through two campaigns: water and oxygen levels in the vacuum vessel are reduced significantly. As a result, KSTAR has achieved its first L-H mode transition, although the input power was marginal for the L-H transition threshold. The characteristics of boron-containing thin films deposited for boronization are investigated.

Wall locking and multiple nonlinear states of magnetic islands

International Nuclear Information System (INIS)

Persson, Mikael; Australian National Univ., Canberra, ACT

1994-01-01

The nonlinear evolution of magnetic islands is analysed in configurations with multiple resonant magnetic surfaces. The existence of multiple nonlinear steady states, is discussed. These are shown to be associated with states where the dynamics around the different rational surfaces are coupled or decoupled and in the presence of a wall of finite resistivity may correspond wall-locked or non-wall-locked magnetic islands. For the case of strong wall stabilization the locking is shown to consist of two different phases. During the first phase the locking of the plasma at the different rational surfaces occurs. Only when the outermost resonant magnetic surface has locked to the inner surfaces can the actual wall locking process take place. Consequently, wall locking, of a global mode, involving more than one rational surface, can be prevented by the decoupling of the resonant magnetic surfaces by plasma rotation. Possible implications on tokamak experiments are discussed. (author)

Domain walls and fermion scattering in grand unified models

International Nuclear Information System (INIS)

Steer, D.A.; Vachaspati, T.

2006-01-01

Motivated by grand unification, we study the properties of domain walls formed in a model with SU(5)xZ 2 symmetry which is spontaneously broken to SU(3)xSU(2)xU(1)/Z 6 , and subsequently to SU(3)xU(1)/Z 3 . Even after the first stage of symmetry breaking, the SU(3) symmetry is broken to SU(2)xU(1)/Z 2 on the domain wall. In a certain range of parameters, flux tubes carrying color- and hyper-charge live on the domain wall and appear as 'boojums' when viewed from one side of the domain wall. Magnetic monopoles are also formed in the symmetry breaking and those carrying color and hyper-charge can be repelled from the wall due to the Meissner effect, or else their magnetic flux can penetrate the domain wall in quantized units. After the second stage of symmetry breaking, fermions can transmute when they scatter with the domain wall, providing a simpler version of fermion-monopole scattering: for example, neutrinos can scatter into d-quarks, leaving behind electric charge and color which is carried by gauge field excitations living on the domain wall

When double-wall carbon nanotubes can become metallic or semiconducting

International Nuclear Information System (INIS)

Moradian, Rostam; Azadi, Sam; Refii-tabar, Hashem

2007-01-01

The electronic properties of double-wall carbon nanotubes (DWCNTs) are investigated via density functional theory. The DWCNTs are separated into four categories wherein the inner-outer nanotubes are metal-metal, metal-semiconductor, semiconductor-metal and semiconductor-semiconductor single-wall nanotubes. The band structure of the DWCNTs, the local density of states of the inner and outer nanotubes, and the total density of states are calculated. We found that for the metal-metal DWCNTs, the inner and outer nanotubes remain metallic for different distances between the walls, while for the metal-semiconductor DWCNTs, decreasing the distance between the walls leads to a phase transition in which both nanotubes become metallic. In the case of semiconductor-metal DWCNTs, it is found that at some distance the inner wall becomes metallic, while the outer wall becomes a semiconductor, and if the distance is decreased, both walls become metallic. Finally, in the semiconductor-semiconductor DWCNTs, if the two walls are far from each other, then the whole DWCNT and both walls remain semiconducting. By decreasing the wall distance, first the inner, and then the outer, nanotube becomes metallic

From Boltzmann equations to steady wall velocities

International Nuclear Information System (INIS)

Konstandin, Thomas; Rues, Ingo; Nardini, Germano; California Univ., Santa Barbara, CA

2014-07-01

By means of a relativistic microscopic approach we calculate the expansion velocity of bubbles generated during a first-order electroweak phase transition. In particular, we use the gradient expansion of the Kadanoff-Baym equations to set up the fluid system. This turns out to be equivalent to the one found in the semi-classical approach in the non-relativistic limit. Finally, by including hydrodynamic deflagration effects and solving the Higgs equations of motion in the fluid, we determine velocity and thickness of the bubble walls. Our findings are compared with phenomenological models of wall velocities. As illustrative examples, we apply these results to three theories providing first-order phase transitions with a particle content in the thermal plasma that resembles the Standard Model.

Changes in Cell Wall Polysaccharides Associated With Growth 1

Science.gov (United States)

Nevins, Donald J.; English, Patricia D.; Albersheim, Peter

1968-01-01

Changes in the polysaccharide composition of Phaseolus vulgaris, P. aureus, and Zea mays cell walls were studied during the first 28 days of seedling development using a gas chromatographic method for the analysis of neutral sugars. Acid hydrolysis of cell wall material from young tissues liberates rhamnose, fucose, arabinose, xylose, mannose, galactose, and glucose which collectively can account for as much as 70% of the dry weight of the wall. Mature walls in fully expanded tissues of these same plants contain less of these constituents (10%-20% of dry wt). Gross differences are observed between developmental patterns of the cell wall in the various parts of a seedling, such as root, stem, and leaf. The general patterns of wall polysaccharide composition change, however, are similar for analogous organs among the varieties of a species. Small but significant differences in the rates of change in sugar composition were detected between varieties of the same species which exhibited different growth patterns. The cell walls of species which are further removed phylogenetically exhibit even more dissimilar developmental patterns. The results demonstrate the dynamic nature of the cell wall during growth as well as the quantitative and qualitative exactness with which the biosynthesis of plant cell walls is regulated. PMID:16656862

The Specific Nature of Plant Cell Wall Polysaccharides 1

Science.gov (United States)

Nevins, Donald J.; English, Patricia D.; Albersheim, Peter

1967-01-01

Polysaccharide compositions of cell walls were assessed by quantitative analyses of the component sugars. Cell walls were hydrolyzed in 2 n trifluoroacetic acid and the liberated sugars reduced to their respective alditols. The alditols were acetylated and the resulting alditol acetates separated by gas chromatography. Quantitative assay of the alditol acetates was accomplished by electronically integrating the detector output of the gas chromatograph. Myo-inositol, introduced into the sample prior to hydrolysis, served as an internal standard. The cell wall polysaccharide compositions of plant varieties within a given species are essentially identical. However, differences in the sugar composition were observed in cell walls prepared from different species of the same as well as of different genera. The fact that the wall compositions of different varieties of the same species are the same indicates that the biosynthesis of cell wall polysaccharides is genetically regulated. The cell walls of various morphological parts (roots, hypocotyls, first internodes and primary leaves) of bean plants were each found to have a characteristic sugar composition. It was found that the cell wall sugar composition of suspension-cultured sycamore cells could be altered by growing the cells on different carbon sources. This demonstrates that the biosynthesis of cell wall polysaccharides can be manipulated without fatal consequences. PMID:16656594

Extended Multiscale Image Segmentation for Castellated Wall Management

Science.gov (United States)

Sakamoto, M.; Tsuguchi, M.; Chhatkuli, S.; Satoh, T.

2018-05-01

Castellated walls are positioned as tangible cultural heritage, which require regular maintenance to preserve their original state. For the demolition and repair work of the castellated wall, it is necessary to identify the individual stones constituting the wall. However, conventional approaches using laser scanning or integrated circuits (IC) tags were very time-consuming and cumbersome. Therefore, we herein propose an efficient approach for castellated wall management based on an extended multiscale image segmentation technique. In this approach, individual stone polygons are extracted from the castellated wall image and are associated with a stone management database. First, to improve the performance of the extraction of individual stone polygons having a convex shape, we developed a new shape criterion named convex hull fitness in the image segmentation process and confirmed its effectiveness. Next, we discussed the stone management database and its beneficial utilization in the repair work of castellated walls. Subsequently, we proposed irregular-shape indexes that are helpful for evaluating the stone shape and the stability of the stone arrangement state in castellated walls. Finally, we demonstrated an application of the proposed method for a typical castellated wall in Japan. Consequently, we confirmed that the stone polygons can be extracted with an acceptable level. Further, the condition of the shapes and the layout of the stones could be visually judged with the proposed irregular-shape indexes.

Fabrication of a 1/6-scale mock-up and manifolds for the Korea first wall in the ITER

International Nuclear Information System (INIS)

Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Lee, Dong Won

2012-01-01

Korea has developed and participated in the Test Blanket Module (TBM) program of the International Thermo-nuclear Experimental Reactor (ITER). The first wall (FW) of the TBM is an important component that faces the plasma directly and therefore it is subjected to high heat and neutron loads. To fabricate the TBM FW, the Hot Isostatic Pressing (HIP) bonding method has been investigated. In the present study, the manufacturing method of the TBM FW is introduced through the fabrication and testing of a 1/6-scale mockup. To distribute fluid uniformly in the mock-up, a manifold was designed and fabricated using the ANSYS-CFX analysis. After the mock-up was fabricated and its fluid distribution tests performed, we compared the results of tests with the simulated results

Conduction at domain walls in oxide multiferroics

Science.gov (United States)

Seidel, J.; Martin, L. W.; He, Q.; Zhan, Q.; Chu, Y.-H.; Rother, A.; Hawkridge, M. E.; Maksymovych, P.; Yu, P.; Gajek, M.; Balke, N.; Kalinin, S. V.; Gemming, S.; Wang, F.; Catalan, G.; Scott, J. F.; Spaldin, N. A.; Orenstein, J.; Ramesh, R.

2009-03-01

Domain walls may play an important role in future electronic devices, given their small size as well as the fact that their location can be controlled. Here, we report the observation of room-temperature electronic conductivity at ferroelectric domain walls in the insulating multiferroic BiFeO3. The origin and nature of the observed conductivity are probed using a combination of conductive atomic force microscopy, high-resolution transmission electron microscopy and first-principles density functional computations. Our analyses indicate that the conductivity correlates with structurally driven changes in both the electrostatic potential and the local electronic structure, which shows a decrease in the bandgap at the domain wall. Additionally, we demonstrate the potential for device applications of such conducting nanoscale features.

Direct numerical simulation of MHD flow with electrically conducting wall

International Nuclear Information System (INIS)

Satake, S.; Kunugi, T.; Naito, N.; Sagara, A.

2006-01-01

The 2D vortex problem and 3D turbulent channel flow are treated numerically to assess the effect of electrically conducting walls on turbulent MHD flow. As a first approximation, the twin vortex pair is considered as a model of a turbulent eddy near the wall. As the eddy approaches and collides with the wall, a high value electrical potential is induced inside the wall. The Lorentz force, associated with the potential distribution, reduces the velocity gradient in the near-wall region. When considering a fully developed turbulent channel flow, a high electrical conductivity wall was chosen to emphasize the effect of electromagnetic coupling between the wall and the flow. The analysis was performed using DNS. The results are compared with a non-MHD flow and MHD flow in the insulated channel. The mean velocity within the logarithmic region in the case of the electrically conducting wall is slightly higher than that in the non-conducting wall case. Thus, the drag is smaller compared to that in the non-conducting wall case due to a reduction of the Reynolds stress in the near wall region through the Lorentz force. This mechanism is explained via reduction of the production term in the Reynolds shear stress budget

Development of a high-quality cut-off wall using electrophoresis

International Nuclear Information System (INIS)

Kawachi, T.; Murahashi, H.

1991-01-01

Techniques to build a high-quality cut-off wall have been developed for storage facilities of low-level radioactive waste (LLW) as an emergency measures to prevent leakages. The cut-off wall is highly impermeable, nucleid-adsorptive and have long-term durability. Electrophoresis is used to form impermeable membrane of bentonite as main features of the cut-off wall. First of all, laboratory tests have been conducted to study ways of building barriers on site and to collect data on the barriers properties. Afterwards, on-site construction tests of a high-quality cut-off wall have been carried out. In this paper, we describe the process and results on the studies of the high-quality cut-off wall using electrophoresis

Behaviour of neutrons passing through the Bloch wall

International Nuclear Information System (INIS)

Schaerpf, O.

1976-01-01

In part I of the present paper the pertinent knowledge about Bloch walls is presented and developed insofar as it appears necessary for the experiments with neutrons, that is to say the direction of magnetization within the domains, the calculation of the variation of magnetization in the wall, the wall thickness, and the zigzag structure of the Bloch wall. In part II it is first clarified why the Bloch wall can be treated as a continuum problem. It shows that this is possible far away from Laue reflexes. For angles far away from Laure-reflex angles the interaction of the periodic structure of the magnetization can be described with the aid of an averaged magnetic flux density. The consequence of it is the possibility of treating the problem by means of a Schroedinger equation with continous interaction. This leads to a law of refraction. The question of the possibilities for explaining the intensity behavior is treated in part III. This part, from different aspects, describes the fact, which already was pointed out in Schaerpf, O., Vehoff, H., Schwink, Ch. 1973, that the spin of the neutrons in passing through the wall is partly taken along by the magnetization gradually rotating in the wall. (orig./WBU) [de

In-Pile thermal fatigue of First Wall mock-ups under ITER relevant conditions

International Nuclear Information System (INIS)

Blom, F.; Schmalz, F.; Kamer, S.; Ketema, D.J.

2006-01-01

The objective of this study is to perform in-pile thermal fatigue testing of three actively cooled First Wall (FW) mock-ups to check the effect of neutron irradiation on the Be/CuCrZr joints under representative FW operation conditions. Three FW mock-ups with Beryllium armor tiles will be neutron irradiated at 1 dpa (in Be) with parallel thermal fatigue testing for 30,000 cycles. The temperatures, stress distributions and stress amplitudes at the Be/CuCrZr interface of the mock-ups will be as close as possible to the values calculated for ITER FW panels. For this objective the PWM mocks-up subjected to thermal fatigue will be integrated with high density (W) plates on the Be-side to provide heat flux by nuclear heating. The assembly will be placed in the pool-side facility of the HFR and thermal cycling is then arranged by mechanical movement towards and from the core box. As the thermal design of the irradiation rig is very critical a pilot-irradiation will be performed to cross check the models used in the thermal design of the rig. The project is currently in the design phase of both the pilot and actual irradiation rig. The irradiation of the actual rig is planned to start at mid 2007 and last for two years. (author)

Effect of activation cross-section uncertainties on the radiological assessment of the MFE/DEMO first wall

Energy Technology Data Exchange (ETDEWEB)

Cabellos, O. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid (Spain)]. E-mail: cabellos@din.upm.es; Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid (Spain); University Nacional Educacion a Distancia, Dep. Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Rodriguez, A. [University Nacional Educacion a Distancia, Dep. Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Youssef, M. [University of California, Los Angeles, CA (United States); Sawan, M. [University of Wisconsin, Madison, WI (United States)

2006-02-15

A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross-sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, {sup 6}Co and {sup 94}Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed.

Effect of activation cross-section uncertainties on the radiological assessment of the MFE/DEMO first wall

International Nuclear Information System (INIS)

Cabellos, O.; Reyes, S.; Sanz, J.; Rodriguez, A.; Youssef, M.; Sawan, M.

2006-01-01

A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross-sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, 6 Co and 94 Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed

First-principles study of structural and work function properties for nitrogen-doped single-walled carbon nanotubes

International Nuclear Information System (INIS)

Shao, Xiji; Li, Detian; Cai, Jianqiu; Luo, Haijun; Dong, Changkun

2016-01-01

Graphical abstract: - Highlights: • Substitutional nitrogen atom doping in capped (5, 5) SWNT is investigated. • Serious defects appear from breaks of C−N bonds with N contents of above 23.3 at.%. • Work function drops after N doping and may reach 4.1 eV. - Abstract: The structural and electronic properties of the capped (5, 5) single-walled carbon nanotube (SWNT), including the structural stability, the work function, and the charge transfer performance, are investigated for the substitutional nitrogen atom doping under different concentrations by first-principles density functional theory. The geometrical structure keeps almost intact with single or two N atom doping, while C−N bonds may break up with serious defects for N concentrations of 23.3 at.% and above. The SWNT remains metallic and the work function drops after doping due to the upward shift of Fermi level, leading to the increase of the electrical conductivity. N doping enhances the oxygen reduction activity stronger than N adsorption because of higher charge transfers.

Improved interior wall detection using designated dictionaries in compressive urban sensing problems

Science.gov (United States)

Lagunas, Eva; Amin, Moeness G.; Ahmad, Fauzia; Nájar, Montse

2013-05-01

In this paper, we address sparsity-based imaging of building interior structures for through-the-wall radar imaging and urban sensing applications. The proposed approach utilizes information about common building construction practices to form an appropriate sparse representation of the building layout. With a ground based SAR system, and considering that interior walls are either parallel or perpendicular to the exterior walls, the antenna at each position would receive reflections from the walls parallel to the radar's scan direction as well as from the corners between two meeting walls. We propose a two-step approach for wall detection and localization. In the first step, a dictionary of possible wall locations is used to recover the positions of both interior and exterior walls that are parallel to the scan direction. A follow-on step uses a dictionary of possible corner reflectors to locate wall-wall junctions along the detected wall segments, thereby determining the true wall extents and detecting walls perpendicular to the scan direction. The utility of the proposed approach is demonstrated using simulated data.

Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

OpenAIRE

Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Kozub, Thomas; Berzak, Laura; Hammett, Gregory; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Jacobson, Craig M.; Lucia, Matthew; Jones, Andrew; Lundberg, Daniel; Timberlake, John; Majeski, Richard

2010-01-01

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas wa...

Radiological diagnosis of chest wall tuberculosis: CT versus chest radiograph

International Nuclear Information System (INIS)

Liu Fugeng; Pan Jishu; Chen Qihang; Zhou Cheng; Yu Jingying; Tang Dairong

2006-01-01

Objective: To evaluate the role of CT or Chest radiograph in diagnosis of chest wall tuberculosis. Methods: The study population included 21 patients with chest wall tuberculosis confirmed by operation or biopsy. Chest radiograph and plain CT were performed in all eases, while enhanced CT in 9 cases, and all images were reviewed by 2 radiologists. Results: Single soft tissue mass of the chest wall was detected in all cases on CT, but not on chest radiograph(χ 2 =42.000, P 2 =4.421, P<0.05). Conclusion: CT, especially enhanced CT scan is the first choice in the diagnosis of chest wall tuberculosis. (authors)

2D-immunoblotting analysis of Sporothrix schenckii cell wall

Directory of Open Access Journals (Sweden)

Estela Ruiz-Baca

2011-03-01

Full Text Available We utilized two-dimensional gel electrophoresis and immunoblotting (2D-immunoblotting with anti-Sporothrix schenckii antibodies to identify antigenic proteins in cell wall preparations obtained from the mycelial and yeast-like morphologies of the fungus. Results showed that a 70-kDa glycoprotein (Gp70 was the major antigen detected in the cell wall of both morphologies and that a 60-kDa glycoprotein was present only in yeast-like cells. In addition to the Gp70, the wall from filament cells showed four proteins with molecular weights of 48, 55, 66 and 67 kDa, some of which exhibited several isoforms. To our knowledge, this is the first 2D-immunoblotting analysis of the S. schenckii cell wall.

Divertor particle exhaust and wall inventory on DIII-D

International Nuclear Information System (INIS)

Maingi, R.; Jackson, G.L.; Mahdavi, M.A.; Schaffer, M.J.; Wade, M.R.; Mioduszewski, P.K.; Hogan, J.T.; Klepper, C.C.; Haas, G.

1995-01-01

Many tokamaks achieve optimum plasma performance by achieving low recycling; various wall conditioning techniques including helium glow discharge cleaning (HeGDC) are routinely applied to help achieve low recycling. Many of these techniques allow strong, transient wall pumping, but they may not be effective for long-pulse tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), the Tokamak Physics Experiment (TPX), Tore Supra Continu, and JT-60SU. Continuous particle exhaust using an in-situ pumping scheme may be effective for wall inventory control in such devices. Recent particle balance experiments on the Tore Supra and DIII-D tokamaks demonstrated that the wall particle inventory could be reduced during a given discharge by use of continuous particle exhaust. In this paper the authors report the first results of wall inventory control and good performance with the in-situ DIII-D cryopump, replacing the HeGDC normally applied between discharges

Post wall fixation by lag screw only in associated both column fractures with posterior wall involvement.

Science.gov (United States)

Wang, Hu; Utku, Kandemir; Zhuang, Yan; Zhang, Kun; Fu, Ya-Hui; Wei, Xing; Wang, Peng-Fei; Cong, Yu-Xuan; Lei, Jin-Lai; Zhang, Bin-Fei

2017-07-01

To evaluate the quality of reduction, clinical outcomes and complications of associated both column acetabular fractures with posterior wall involvement that are treated through single ilioinguinal approach and fixation of posterior wall by lag screws only. We conducted a retrospective review involving ninety-nine consecutive patients with associated both column fractures of acetabulum treated through single ilioinguinal approach. Patients were divided into two groups. The first group consisted of 35 patients presented with both column fractures with posterior wall involvement that fixation performed with lag screws. This group was compared to a second group of 64 patients with both column fractures without posterior wall involvement. The quality of reduction was assessed using criteria described by Matta. The size of posterior wall fragment was measured. Functional outcome was evaluated using Modified Postel Merle D'Aubigne score. Radiographs at the latest follow up were analyzed for arthritis (Kellgren-Lawrence classification), and femoral head avascular necrosis (Ficat/Arlet classification). The study showed no significant differences in all preoperative variables (P>0.05). While intraoperative blood loss and operative time in group 1 were increased compared to group 2, the difference was not statistically significant (P>0.05). The height, relative depth and peripheral length of posterior wall respectively were 27.8±2.5mm (range: 24-35mm), 71.5±5.4% (range: 65-88%), 23.0±2.3mm (range: 17-28mm). The mean posterior wall fracture displacement is 5.0±3.2mm (range: 0-11mm). There was no difference regarding the quality of reduction between the two groups (P>0.05). The excellent to good clinical outcome was around 71.4% in the group 1 versus 73.4% in the group 2 at the final follow-up, this difference was not statistically significant (P>0.05). There was no difference in rate of complications between the two groups (P>0.05). Lag screws fixation of posterior wall

Seismic behavior and design of wall-EDD-frame systems

Directory of Open Access Journals (Sweden)

Oren eLavan

2015-06-01

Full Text Available Walls and frames have different deflection lines and, depending on the seismic mass they support, may often poses different natural periods. In many cases, wall-frame structures present an advantageous behavior. In these structures the walls and the frames are rigidly connected. Nevertheless, if the walls and the frames were not rigidly connected, an opportunity for an efficient passive control strategy would arise: Connecting the two systems by energy dissipation devices (EDDs to result in wall-EDD-frame systems. This, depending on the parameters of the system, is expected to lead to an efficient energy dissipation mechanism.This paper studies the seismic behavior of wall-EDD-frame systems in the context of retrofitting existing frame structures. The controlling non-dimensional parameters of such systems are first identified. This is followed by a rigorous and extensive parametric study that reveals the pros and cons of the new system versus wall-frame systems. The effect of the controlling parameters on the behavior of the new system are analyzed and discussed. Finally, tools are given for initial design of such retrofitting schemes. These enable both choosing the most appropriate retrofitting alternative and selecting initial values for its parameters.

Streaming vorticity flux from oscillating walls with finite amplitude

Science.gov (United States)

Wu, J. Z.; Wu, X. H.; Wu, J. M.

1993-01-01

How to describe vorticity creation from a moving wall is a long standing problem. This paper discusses relevant issues at the fundamental level. First, it is shown that the concept of 'vorticity flux due to wall acceleration' can be best understood by following fluid particles on the wall rather than observing the flow at fixed spatial points. This is of crucial importance when the time-averaged flux is to be considered. The averaged flux has to be estimated in a wall-fixed frame of reference (in which there is no flux due to wall acceleration at all); or, if an inertial frame of reference is used, the generalized Lagrangian mean (GLM) also gives the same result. Then, for some simple but typical configurations, the time-averaged vorticity flux from a harmonically oscillating wall with finite amplitude is analyzed, without appealing to small perturbation. The main conclusion is that the wall oscillation will produce an additional mean vorticity flux (a fully nonlinear streaming effect), which is partially responsible for the mechanism of vortex flow control by waves. The results provide qualitative explanation for some experimentally and/or computationally observed phenomena.

Toward Tungsten Plasma-Facing Components in KSTAR: Research on Plasma-Metal Wall Interaction

NARCIS (Netherlands)

Hong, S. H.; Kim, K. M.; Song, J. H.; Bang, E. N.; Kim, H. T.; Lee, K. S.; Litnovsky, A.; Hellwig, M.; Seo, D. C.; van den Berg, M. A.; Lee, H. H.; Kang, C. S.; Lee, H. Y.; Hong, J. H.; Bak, J. G.; Kim, H. S.; Juhn, J. W.; Son, S. H.; Kim, H. K.; Douai, D.; Grisolia, C.; Wu, J.; Luo, G. N.; Choe, W. H.; Komm, M.; De Temmerman, G.; Pitts, R.

2015-01-01

One of the main missions of KSTAR is to develop long-pulse operation capability relevant to the production of fusion energy. After a full metal wall configuration was decided for ITER, a major upgrade for KSTAR was planned, to a tungsten first wall similar to the JET ITER-like wall (coatings and

Wall-based identification of coherent structures in wall-bounded turbulence

Science.gov (United States)

Sanmiguel Vila, C.; Flores, O.

2018-04-01

During the last decades, a number of reduced order models based on coherent structures have been proposed to describe wall-bounded turbulence. Many of these models emphasize the importance of coherent wall-normal velocity eddies (ν-eddies), which drive the generation of the very long streamwise velocity structures observed in the logarithmic and outer region. In order to use these models to improve our ability to control wall-bounded turbulence in realistic applications, these ν-eddies need to be identified from the wall in a non-intrusive way. In this paper, the possibility of using the pressure signal at the wall to identify these ν-eddies is explored, analyzing the cross-correlation between the wall-normal velocity component and the pressure fluctuations at the wall in a DNS of a turbulent channel flow at Reτ = 939. The results show that the cross-correlation has a region of negative correlation upstream, and a region of positive correlation backwards. In the spanwise direction the correlation decays monotonously, except very close to the wall where a change of sign of the correlation coefficient is observed. Moreover, filtering the pressure fluctuations at the wall in space results in an increase of the region where the cross-correlation is strong, both for the positively and the negatively correlated regions. The use of a time filter for the pressure fluctuations at the wall yields different results, displacing the regions of strong correlation without changing much their sizes. The results suggest that space-filtering the pressure at the wall is a feasible way to identify ν-eddies of different sizes, which could be used to trigger turbulent control strategies.

Skin effect modifications of the Resistive Wall Mode dynamics in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Villone, Fabio, E-mail: villone@unicas.it [Ass. Euratom/ENEA/CREATE, DIEI, Università di Cassino e del Lazio Meridionale, Via Di Biasio 43, 03043 Cassino, FR (Italy); Pustovitov, Vladimir D. [Institute of Tokamak Physics, National Research Centre ‘Kurchatov Institute’, Pl. Kurchatova 1, Moscow 123182 (Russian Federation)

2013-11-22

We present the first evidence of the skin-effect modification of the Resistive Wall Mode (RWM) dynamics in a tokamak. The computations are performed with the CarMa code, using its unique ability of treating volumetric 3D conducting structures. The results prove that conventional thin-wall models and codes, assuming the thin equivalent wall located on the inner side of a real (thick) wall, may fail to get accurate estimates of RWM growth rates, since the inclusion of the skin effect makes the growth rates always larger than otherwise. The difference is noticeable even for the conventional slow RWMs and becomes substantial for faster modes. Some possible equivalent thin-wall modeling approaches are also discussed.

FRP strengthening of RC walls with openings

DEFF Research Database (Denmark)

Hansen, Christian Skodborg; Sas, Gabriel; Täljsten, Björn

2009-01-01

Strengthening reinforced concrete (RC) walls with openings using fibre reinforced polymers (FRP) has been experimentally proven to be a viable rehabilitation method. However, very few theoretical investigations are reported. In this paper two methods of analysis are presented. Since openings vary...... in size, the analysis of a strengthened wall can be divided into frame idealization method for large openings, and combined disk and frame analysis for smaller openings. The first method provides an easy to use tool in practical engineering, where the latter describes the principles of a ductile...

Conceptual design of a First Wall mock-up experiment in preparation for the qualification of breeding blanket technologies in the Helium Loop Karlsruhe (HELOKA) facility

Energy Technology Data Exchange (ETDEWEB)

Zeile, C., E-mail: christian.zeile@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Abou-Sena, A.; Boccaccini, L.V.; Ghidersa, B.E.; Kang, Q.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Lamberti, L. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Maione, I.A.; Rey, J.; Weth, A. von der [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2016-11-01

Highlights: • Experiment in preparation for the qualification of Breeding Blanket technologies in HELOKA facility is proposed. • Experimental capabilities, instrumentation of the mock-up and experimental program are presented. • Design and manufacturing of the mock-up is described. • Design of modular attachment system to obtain different stress levels and distributions on the mock-up is discussed. - Abstract: An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface.

Using digital holographic microscopy for simultaneous measurements of 3D near wall velocity and wall shear stress in a turbulent boundary layer

Science.gov (United States)

Sheng, J.; Malkiel, E.; Katz, J.

2008-12-01

A digital holographic microscope is used to simultaneously measure the instantaneous 3D flow structure in the inner part of a turbulent boundary layer over a smooth wall, and the spatial distribution of wall shear stresses. The measurements are performed in a fully developed turbulent channel flow within square duct, at a moderately high Reynolds number. The sample volume size is 90 × 145 × 90 wall units, and the spatial resolution of the measurements is 3 8 wall units in streamwise and spanwise directions and one wall unit in the wall-normal direction. The paper describes the data acquisition and analysis procedures, including the particle tracking method and associated method for matching of particle pairs. The uncertainty in velocity is estimated to be better than 1 mm/s, less than 0.05% of the free stream velocity, by comparing the statistics of the normalized velocity divergence to divergence obtained by randomly adding an error of 1 mm/s to the data. Spatial distributions of wall shear stresses are approximated with the least square fit of velocity measurements in the viscous sublayer. Mean flow profiles and statistics of velocity fluctuations agree very well with expectations. Joint probability density distributions of instantaneous spanwise and streamwise wall shear stresses demonstrate the significance of near-wall coherent structures. The near wall 3D flow structures are classified into three groups, the first containing a pair of counter-rotating, quasi streamwise vortices and high streak-like shear stresses; the second group is characterized by multiple streamwise vortices and little variations in wall stress; and the third group has no buffer layer structures.

Root Growth and Water distribution in living walls

DEFF Research Database (Denmark)

Jørgensen, Lars

of functional living walls and this thesis is a first step of understanding the essential but hidden part inside the growing medium, i.e. the roots. Ensuring successful performance of the plants in a living wall is complex and the choice of growing medium, plant species and planting position are important....... for root growth. This thesis investigates the correlations between the growing media and root and shoots growth, and studies root growth patterns of different plant species and effects of planting position and root interactions of plants growing in living walls. There are a number of challenges with living...... walls; the vertical orientation of the growing medium, plants are growing vertically above or below each other in a limited rooting volume; there is an increased exposure to weather and the plants can react differently to water conditions and competition from other plants. Plant growth is the core...

Chalcone Synthase (CHS) Gene Suppression in Flax Leads to Changes in Wall Synthesis and Sensing Genes, Cell Wall Chemistry and Stem Morphology Parameters

Science.gov (United States)

Zuk, Magdalena; Działo, Magdalena; Richter, Dorota; Dymińska, Lucyna; Matuła, Jan; Kotecki, Andrzej; Hanuza, Jerzy; Szopa, Jan

2016-01-01

The chalcone synthase (CHS) gene controls the first step in the flavonoid biosynthesis. In flax, CHS down-regulation resulted in tannin accumulation and reduction in lignin synthesis, but plant growth was not affected. This suggests that lignin content and thus cell wall characteristics might be modulated through CHS activity. This study investigated the possibility that CHS affects cell wall sensing as well as polymer content and arrangement. CHS-suppressed and thus lignin-reduced plants showed significant changes in expression of genes involved in both synthesis of components and cell wall sensing. This was accompanied by increased levels of cellulose and hemicellulose. CHS-reduced flax also showed significant changes in morphology and arrangement of the cell wall. The stem tissue layers were enlarged averagely twofold compared to the control, and the number of fiber cells more than doubled. The stem morphology changes were accompanied by reduction of the crystallinity index of the cell wall. CHS silencing induces a signal transduction cascade that leads to modification of plant metabolism in a wide range and thus cell wall structure. PMID:27446124

Structural domain walls in polar hexagonal manganites

Science.gov (United States)

Kumagai, Yu

2014-03-01

The domain structure in the multiferroic hexagonal manganites is currently intensely investigated, motivated by the observation of intriguing sixfold topological defects at their meeting points [Choi, T. et al,. Nature Mater. 9, 253 (2010).] and nanoscale electrical conductivity at the domain walls [Wu, W. et al., Phys. Rev. Lett. 108, 077203 (2012).; Meier, D. et al., Nature Mater. 11, 284 (2012).], as well as reports of coupling between ferroelectricity, magnetism and structural antiphase domains [Geng, Y. et al., Nano Lett. 12, 6055 (2012).]. The detailed structure of the domain walls, as well as the origin of such couplings, however, was previously not fully understood. In the present study, we have used first-principles density functional theory to calculate the structure and properties of the low-energy structural domain walls in the hexagonal manganites [Kumagai, Y. and Spaldin, N. A., Nature Commun. 4, 1540 (2013).]. We find that the lowest energy domain walls are atomically sharp, with {210}orientation, explaining the orientation of recently observed stripe domains and suggesting their topological protection [Chae, S. C. et al., Phys. Rev. Lett. 108, 167603 (2012).]. We also explain why ferroelectric domain walls are always simultaneously antiphase walls, propose a mechanism for ferroelectric switching through domain-wall motion, and suggest an atomistic structure for the cores of the sixfold topological defects. This work was supported by ETH Zurich, the European Research Council FP7 Advanced Grants program me (grant number 291151), the JSPS Postdoctoral Fellowships for Research Abroad, and the MEXT Elements Strategy Initiative to Form Core Research Center TIES.

Wall Shear Stress, Wall Pressure and Near Wall Velocity Field Relationships in a Whirling Annular Seal

Science.gov (United States)

Morrison, Gerald L.; Winslow, Robert B.; Thames, H. Davis, III

1996-01-01

The mean and phase averaged pressure and wall shear stress distributions were measured on the stator wall of a 50% eccentric annular seal which was whirling in a circular orbit at the same speed as the shaft rotation. The shear stresses were measured using flush mounted hot-film probes. Four different operating conditions were considered consisting of Reynolds numbers of 12,000 and 24,000 and Taylor numbers of 3,300 and 6,600. At each of the operating conditions the axial distribution (from Z/L = -0.2 to 1.2) of the mean pressure, shear stress magnitude, and shear stress direction on the stator wall were measured. Also measured were the phase averaged pressure and shear stress. These data were combined to calculate the force distributions along the seal length. Integration of the force distributions result in the net forces and moments generated by the pressure and shear stresses. The flow field inside the seal operating at a Reynolds number of 24,000 and a Taylor number of 6,600 has been measured using a 3-D laser Doppler anemometer system. Phase averaged wall pressure and wall shear stress are presented along with phase averaged mean velocity and turbulence kinetic energy distributions located 0.16c from the stator wall where c is the seal clearance. The relationships between the velocity, turbulence, wall pressure and wall shear stress are very complex and do not follow simple bulk flow predictions.

Increased Bladder Wall Thickness in Diabetic and Nondiabetic Women With Overactive Bladder

Directory of Open Access Journals (Sweden)

Hakkı Uzun

2013-06-01

Full Text Available Purpose: Bladder wall thickness has been reported to be associated with overactive bladder (OAB in women. Diabetic women have an increased risk for OAB syndrome and may have an increased risk for bladder wall thickness. Methods: A total of 235 female patients aged 40 to 75 years were categorized into four groups. The first group consisted of women free of urgency or urge urinary incontinence. The second group included nondiabetic women with idiopathic OAB. The third group consisted of women with diabetes and clinical OAB, and women with diabetes but without OAB constituted the fourth group. Bladder wall thickness at the anterior wall was measured by ultrasound by the suprapubic approach with bladder filling over 250 mL. Results: The diabetic (third group and nondiabetic (second group women with OAB had significantly greater bladder wall thickness at the anterior bladder wall than did the controls. However, the difference was not significant between the diabetic (third group and the nondiabetic (second group women with OAB. Women with diabetes but without OAB (fourth group had greater bladder wall thickness than did the controls but this difference was not significant. Additionally, the difference in bladder wall thickness between diabetic women with (third group and without (fourth group OAB was not significant. Conclusions: This is the first study to show that bladder wall thickness is increased in diabetic women with and without OAB. Additionally, nondiabetic women with OAB had increased bladder wall thickness. Further studies may provide additional information for diabetic and nondiabetic women with OAB, in whom the etiopathogenesis of the disease may be similar.

Ambiguous walls

DEFF Research Database (Denmark)

Mody, Astrid

2012-01-01

The introduction of Light Emitting Diodes (LEDs) in the built environment has encouraged myriad applications, often embedded in surfaces as an integrated part of the architecture. Thus the wall as responsive luminous skin is becoming, if not common, at least familiar. Taking into account how walls...... have encouraged architectural thinking of enclosure, materiality, construction and inhabitation in architectural history, the paper’s aim is to define new directions for the integration of LEDs in walls, challenging the thinking of inhabitation and program. This paper introduces the notion...... of “ambiguous walls” as a more “critical” approach to design [1]. The concept of ambiguous walls refers to the diffuse status a lumious and possibly responsive wall will have. Instead of confining it can open up. Instead of having a static appearance, it becomes a context over time. Instead of being hard...

Resonant tunneling across a ferroelectric domain wall

Science.gov (United States)

Li, M.; Tao, L. L.; Velev, J. P.; Tsymbal, E. Y.

2018-04-01

Motivated by recent experimental observations, we explore electron transport properties of a ferroelectric tunnel junction (FTJ) with an embedded head-to-head ferroelectric domain wall, using first-principles density-functional theory calculations. We consider a FTJ with L a0.5S r0.5Mn O3 electrodes separated by a BaTi O3 barrier layer and show that an in-plane charged domain wall in the ferroelectric BaTi O3 can be induced by polar interfaces. The resulting V -shaped electrostatic potential profile across the BaTi O3 layer creates a quantum well and leads to the formation of a two-dimensional electron gas, which stabilizes the domain wall. The confined electronic states in the barrier are responsible for resonant tunneling as is evident from our quantum-transport calculations. We find that the resonant tunneling is an orbital selective process, which leads to sharp spikes in the momentum- and energy-resolved transmission spectra. Our results indicate that domain walls embedded in FTJs can be used to control the electron transport.

The dorsal shell wall structure of Mesozoic ammonoids

Directory of Open Access Journals (Sweden)

Gregor Radtke

2017-03-01

Full Text Available The study of pristine preserved shells of Mesozoic Ammonoidea shows different types of construction and formation of the dorsal shell wall. We observe three major types: (i The vast majority of Ammonoidea, usually planispirally coiled, has a prismatic reduced dorsal shell wall which consists of an outer organic component (e.g., wrinkle layer, which is the first layer to be formed, and the subsequently formed dorsal inner prismatic layer. The dorsal mantle tissue suppresses the formation of the outer prismatic layer and nacreous layer. With the exception of the outer organic component, secretion of a shell wall is omitted at the aperture. A prismatic reduced dorsal shell wall is always secreted immediately after the hatching during early teleoconch formation. Due to its broad distribution in (planispiral Ammonoidea, the prismatic reduced dorsal shell wall is probably the general state. (ii Some planispirally coiled Ammonoidea have a nacreous reduced dorsal shell wall which consists of three mineralized layers: two prismatic layers (primary and secondary dorsal inner prismatic layer and an enclosed nacreous layer (secondary dorsal nacreous layer. The dorsal shell wall is omitted at the aperture and was secreted in the rear living chamber. Its layers are a continuation of an umbilical shell doubling (reinforcement by additional shell layers that extends towards the ventral crest of the preceding whorl. The nacreous reduced dorsal shell wall is formed in the process of ontogeny following a prismatic reduced dorsal shell wall. (iii Heteromorph and some planispirally coiled taxa secrete a complete dorsal shell wall which forms a continuation of the ventral and lateral shell layers. It is formed during ontogeny following a prismatic reduced dorsal shell wall or a priori. The construction is identical with the ventral and lateral shell wall, including a dorsal nacreous layer. The wide distribution of the ability to form dorsal nacre indicates that it is

Powder metallurgical processing of self-passivating tungsten alloys for fusion first wall application

International Nuclear Information System (INIS)

López-Ruiz, P.; Ordás, N.; Iturriza, I.; Walter, M.; Gaganidze, E.; Lindig, S.; Koch, F.; García-Rosales, C.

2013-01-01

Self-passivating tungsten based alloys are expected to provide a major safety advantage compared to pure tungsten, presently the main candidate material for first wall armour of future fusion reactors. In case of a loss of coolant accident with simultaneous air ingress, a protective oxide scale will be formed on the surface of W avoiding the formation of volatile and radioactive WO 3 . Bulk WCr12Ti2.5 alloys were manufactured by mechanical alloying (MA) and hot isostatic pressing (HIP), and their properties compared to bulk WCr10Si10 alloys from previous work. The MA parameters were adjusted to obtain the best balance between lowest possible amount of contaminants and effective alloying of the elemental powders. After HIP, a density >99% is achieved for the WCr12Ti2.5 alloy and a very fine and homogeneous microstructure with grains in the submicron range is obtained. Unlike the WCr10Si10 material, no intergranular ODS phase inhibiting grain growth was detected. The thermal and mechanical properties of the WCr10Si10 material are dominated by the silicide (W,Cr) 5 Si 3 ; it shows a sharp ductile-to brittle transition in the range 1273–1323 K. The thermal conductivity of the WCr12Ti2.5 alloy is close to 50 W/mK in the temperature range of operation; it exhibits significantly higher strength and lower DBTT – around 1170 K – than the WCr10Si10 material

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

International Nuclear Information System (INIS)

Wong, C.P.C.; Malang, S.; Sawan, M.

2005-01-01

As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R and D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan

Far SOL transport and main wall plasma interaction in DIII-D

International Nuclear Information System (INIS)

Rudakov, D.L.; Boedo, J.A.; Moyer, R.A.; Doerner, R.P.; Hollmann, E.M.; Krasheninnikov, S.I.; Pigarov, A.Yu.; Stangeby, P.C.; McLean, A.G.; Watkins, J.G.; Wampler, W.R.; Whyte, D.G.; McKee, G.R.; Zeng, L.; Wang, G.; Brooks, N.H.; Evans, T.E.; Leonard, A.W.; Mahdavi, M.A.; West, W.P.; Wong, C.P.C.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.

2005-01-01

Far scrape-off layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma-wall contact. In H-mode between edge localized modes (ELMs), plasma-wall contact is generally weaker than in L-mode. During ELMs plasma fluxes to the wall increase to, or above the L-mode levels. Depending on the discharge conditions ELMs are responsible for 30-90% of the ion flux to the outboard chamber wall. Cross-field fluxes in far SOL are dominated by large amplitude intermittent transport events that may propagate all the way to the outer wall and cause sputtering. A Divertor Material Evaluation System (DiMES) probe containing samples of several ITER-relevant materials including carbon, beryllium and tungsten was exposed to a series of upper single null (USN) discharges as a proxy to measure the first wall erosion. (author)

Advanced Extended Plate and Beam Wall System in a Cold-Climate House

Energy Technology Data Exchange (ETDEWEB)

Mallay, Dave [Partnership for Home Innovation, Upper Marlboro, MD (United States); Wiehagen, Joseph [Partnership for Home Innovation, Upper Marlboro, MD (United States); Kochkin, Vladimir [Partnership for Home Innovation, Upper Marlboro, MD (United States)

2016-01-01

This report presents the design and evaluation of an innovative wall system. This highly insulated (high-R) light-frame wall system for use above grade in residential buildings is referred to as Extended Plate & Beam (EP&B). The EP&B design is the first of its kind to be featured in a new construction test house (NCTH) for the DOE Building America program. The EP&B wall design integrates standard building methods and common building products to construct a high-R wall that minimizes transition risks and costs to builders.

High-R Walls for Remodeling: Wall Cavity Moisture Monitoring

Energy Technology Data Exchange (ETDEWEB)

Wiehagen, J.; Kochkin, V.

2012-12-01

The focus of the study is on the performance of wall systems, and in particular, the moisture characteristics inside the wall cavity and in the wood sheathing. Furthermore, while this research will initially address new home construction, the goal is to address potential moisture issues in wall cavities of existing homes when insulation and air sealing improvements are made.

High-R Walls for Remodeling. Wall Cavity Moisture Monitoring

Energy Technology Data Exchange (ETDEWEB)

Wiehagen, J. [NAHB Research Center Industry Partnership, Upper Marlboro, MD (United States); Kochkin, V. [NAHB Research Center Industry Partnership, Upper Marlboro, MD (United States)

2012-12-01

The focus of the study is on the performance of wall systems, and in particular, the moisture characteristics inside the wall cavity and in the wood sheathing. Furthermore, while this research will initially address new home construction, the goal is to address potential moisture issues in wall cavities of existing homes when insulation and air sealing improvements are made.

Summary report of the IAEA advisory group meeting on nuclear data for neutron multiplication in fusion-reactor first-wall and blanket materials

International Nuclear Information System (INIS)

Muir, D.W.; Pashchenko, A.B.

1992-09-01

The present Report contains the Summary of the IAEA Advisory Group Meeting on Nuclear Data for Neutron Multiplication in Fusion-Reactor First-Wall and Blanket Materials, which was hosted by the Southwest Institute of Nuclear physics and Chemistry (SWINPC) at Chengdu, China and held from 19-21 November 1990. This AGM was organized by the IAEA Nuclear Data Section (NDS), with the cooperation and assistance of local organizers at the SWINPC. The papers which the participants prepared for and presented at the meeting will be published as an INDC report. (author)

Plasma-wall interaction in NET

International Nuclear Information System (INIS)

Engelmann, F.; Chazalon, M.; Moons, F.; Vieider, G.; Harrison, M.F.A.; Hotston, E.S.

1987-01-01

NET is conceived as an experimental reactor with the aim of demonstrating reactor-relevant plasma performance and reliable operation of the device as well as developing and testing components for a demonstration reactor. For power and particle exhaust both a single-null and a double-null poloidal divertor configuration are under consideration. An intense modelling effort is undertaken to predict the heat load and erosion characteristics for these configurations. Under burn conditions, the divertor will operate in the high-recycling regime. The resulting heat loads on the divertor plates are predicted to be somewhat more demanding in the case of a single-null divertor. If one excludes working under conditions where a large part of the power is exhausted by radiation from the plasma edge, refractory metals (W, Mo) have to be used for the plasma-facing surface of the divertor plates, the radial heat and particle transport in the scrape-off layer must be large and the plasma density at the edge of the discharge must be high (n s ≅ 5x10 19 m -3 ). Erosion of a bare stainless steel first wall, under normal working conditions, appears to be within acceptable limits, but the use of graphite armouring is considered in order to avoid wall damage due to localized loads of highly energetic particles and to protect against disruption. Such a solution would also be consistent with the anticipated requirements during start-up. For both the first wall and the divertor plates various concepts are under consideration. Using replaceable tiles as plasma-facing components throughout appears attractive. (orig./GG)

Wall shear stress fixed points in blood flow

Science.gov (United States)

Arzani, Amirhossein; Shadden, Shawn

2017-11-01

Patient-specific computational fluid dynamics produces large datasets, and wall shear stress (WSS) is one of the most important parameters due to its close connection with the biological processes at the wall. While some studies have investigated WSS vectorial features, the WSS fixed points have not received much attention. In this talk, we will discuss the importance of WSS fixed points from three viewpoints. First, we will review how WSS fixed points relate to the flow physics away from the wall. Second, we will discuss how certain types of WSS fixed points lead to high biochemical surface concentration in cardiovascular mass transport problems. Finally, we will introduce a new measure to track the exposure of endothelial cells to WSS fixed points.

Rising damp in building walls: the wall base ventilation system

Energy Technology Data Exchange (ETDEWEB)

Guimaraes, A.S.; Delgado, J.M.P.Q.; Freitas, V.P. de [Faculdade de Engenharia da Universidade do Porto, Laboratorio de Fisica das Construcoes (LFC), Departamento de Engenharia Civil, Porto (Portugal)

2012-12-15

This work intends to validate a new system for treating rising damp in historic buildings walls. The results of laboratory experiments show that an efficient way of treating rising damp is by ventilating the wall base, using the HUMIVENT technique. The analytical model presented describes very well the observed features of rising damp in walls, verified by laboratory tests, who contributed for a simple sizing of the wall base ventilation system that will be implemented in historic buildings. (orig.)

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

Science.gov (United States)

Deepak, SHARMA; Paritosh, CHAUDHURI

2018-04-01

The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

The Dynamic Similitude Design Method of Thin Walled Structures and Experimental Validation

Directory of Open Access Journals (Sweden)

Zhong Luo

2016-01-01

Full Text Available For the applicability of dynamic similitude models of thin walled structures, such as engine blades, turbine discs, and cylindrical shells, the dynamic similitude design of typical thin walled structures is investigated. The governing equation of typical thin walled structures is firstly unified, which guides to establishing dynamic scaling laws of typical thin walled structures. Based on the governing equation, geometrically complete scaling law of the typical thin walled structure is derived. In order to determine accurate distorted scaling laws of typical thin walled structures, three principles are proposed and theoretically proved by combining the sensitivity analysis and governing equation. Taking the thin walled annular plate as an example, geometrically complete and distorted scaling laws can be obtained based on the principles of determining dynamic scaling laws. Furthermore, the previous five orders’ accurate distorted scaling laws of thin walled annular plates are presented and numerically validated. Finally, the effectiveness of the similitude design method is validated by experimental annular plates.

Continuously renewed wall for a thermonuclear reactor

International Nuclear Information System (INIS)

Livshits, A.I.; Pustovojt, YU.M.; Samartsev, A.A.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii)

1982-01-01

The possibility of creating a continuously renewed first wall of a thermonuclear reactor is experimentally investigated. The following variants of the wall are considered: the wall is double, its part turned to plasma is made of comparatively thin material. The external part separated from it by a small gap appears to be protected from interaction with plasma and performs structural functions. The gap contains the mixture of light helium and hydrogen and carbon-containing gas. The light gas transfers heat from internal part of the wall to the external part. Carbon-containing gas provides continuous renewal of carbon coating of the operating surface. The experiment is performed with palladium membrane 20 μm thick. Carbon is introduced into the membrane by benzol pyrolysis on one of the surfaces at the membrane temperature of 900 K. Carbon removal from the operating side of the wall due to its spraying by fast particles is modelled by chemical itching with oxygen given to the operating membrane wall. Observation of the carbon release on the operating surface is performed mass-spectrometrically according to the observation over O 2 transformation into CO and CO 2 . It is shown that in cases of benzol pressure of 5x10 -7 torr, carbon current on the opposite surface is not less than 3x10 12 atoms/sm 2 s and corresponds to the expected wall spraying rate in CF thermonuclear reactors. It is also shown that under definite conditions the formation and maintaining of a through protective carbon coating in the form of a monolayer or volumetric phase is possible

The Cell Wall of Bacillus subtilis

NARCIS (Netherlands)

Scheffers, Dirk-Jan; Graumann, Peter

2012-01-01

The cell wall of Bacillus subtilis is a rigid structure on the outside of the cell that forms the first barrier between the bacterium and the environment, and at the same time maintains cell shape and withstands the pressure generated by the cell’s turgor. In this chapter, the chemical composition

Localization of vector field on dynamical domain wall

Directory of Open Access Journals (Sweden)

Masafumi Higuchi

2017-03-01

Full Text Available In the previous works (arXiv:1202.5375 and arXiv:1402.1346, the dynamical domain wall, where the four dimensional FRW universe is embedded in the five dimensional space–time, has been realized by using two scalar fields. In this paper, we consider the localization of vector field in three formulations. The first formulation was investigated in the previous paper (arXiv:1510.01099 for the U(1 gauge field. In the second formulation, we investigate the Dvali–Shifman mechanism (arXiv:hep-th/9612128, where the non-abelian gauge field is confined in the bulk but the gauge symmetry is spontaneously broken on the domain wall. In the third formulation, we investigate the Kaluza–Klein modes coming from the five dimensional graviton. In the Randall–Sundrum model, the graviton was localized on the brane. We show that the (5,μ components (μ=0,1,2,3 of the graviton are also localized on the domain wall and can be regarded as the vector field on the domain wall. There are, however, some corrections coming from the bulk extra dimension if the domain wall universe is expanding.

The adventitia: essential regulator of vascular wall structure and function.

Science.gov (United States)

Stenmark, Kurt R; Yeager, Michael E; El Kasmi, Karim C; Nozik-Grayck, Eva; Gerasimovskaya, Evgenia V; Li, Min; Riddle, Suzette R; Frid, Maria G

2013-01-01

The vascular adventitia acts as a biological processing center for the retrieval, integration, storage, and release of key regulators of vessel wall function. It is the most complex compartment of the vessel wall and is composed of a variety of cells, including fibroblasts, immunomodulatory cells (dendritic cells and macrophages), progenitor cells, vasa vasorum endothelial cells and pericytes, and adrenergic nerves. In response to vascular stress or injury, resident adventitial cells are often the first to be activated and reprogrammed to influence the tone and structure of the vessel wall; to initiate and perpetuate chronic vascular inflammation; and to stimulate expansion of the vasa vasorum, which can act as a conduit for continued inflammatory and progenitor cell delivery to the vessel wall. This review presents the current evidence demonstrating that the adventitia acts as a key regulator of vascular wall function and structure from the outside in.

An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

Energy Technology Data Exchange (ETDEWEB)

Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

2006-02-01

An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.